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Sample records for pbmr studi reaktor

  1. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    Directory of Open Access Journals (Sweden)

    Ralind Re Marla

    2015-03-01

    Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

  2. A Study for Burn-up Calculation applied on 400MWth PBMR Core

    International Nuclear Information System (INIS)

    Luu, Nam Hai; Kim, Hong Chul; Kim, Soon Young; Kim, Jong Kyung; Noh, Jae Man

    2007-01-01

    The 400MWth Pebble-bed Modular Reactor (PBMR) is an advanced high temperature gas cooled-reactor (HTGR). It possesses a very high efficiency and attractive economics without compromising the high levels of passive safety expected of advanced nuclear designs. With this reason, PBMR is a target which researchers especially in nuclear engineering field study carefully and therefore it is regarded as the leader in the power generation field. There are many research results about benchmark problems but results of the burn-up process are still poor. Hence, in this study a burn-up calculation was performed with PBMR using MONTEBURNS code in which MCNP modeling linked a depletion systems is used

  3. PBMR desalination options: An economic study - HTR2008-58212

    International Nuclear Information System (INIS)

    De Bruyn, R.; Van Ravenswaay, J. P.; Hannink, R.; Kuhr, R.; Bhagat, K.; Zervos, N.

    2008-01-01

    The Pebble Bed Modular Reactor (PBMR), under development in South Africa, is an advanced helium-cooled graphite moderated high-temperature gas-cooled nuclear reactor. The heat output of the PBMR is primarily suited for process applications or power generation. In addition, various desalination technologies can be coupled to the PBMR to further improve the overall efficiency and economics, where suitable site opportunities exist. Several desalination application concepts were evaluated for both a cogeneration configuration as well as a waste heat utilization configuration. These options were evaluated to compare the relative economics of the different concepts and to determine the feasibility of each configuration. The cogeneration desalination configuration included multiple PBMR units producing steam for a power cycle, using a back-pressure steam turbine generator exhausting into different thermal desalination technologies. These technologies include Multi-Effect Distillation (MED), Multi-Effect Distillation with Thermal Vapor Compression (MED-TVC) as well as Multi-Stage Flash (MSF) with all making use of extraction steam from back-pressure turbines. These configurations are optimized to maximize gross revenue from combined power and desalinated water sales using representative economic assumptions with a sensitivity analysis to observe the impact of varying power and water costs. Increasing turbine back pressure results in a loss of power output but a gain in water production. The desalination systems are compared as incremental investments. A standard MED process with minimal effects appears most attractive, although results are very sensitive with regards to projected power and water values. (authors)

  4. ZZ-PBMR-400, OECD/NEA PBMR Coupled Neutronics/Thermal Hydraulics Transient Benchmark - The PBMR-400 Core Design

    International Nuclear Information System (INIS)

    Reitsma, Frederik

    2007-01-01

    Description of benchmark: This international benchmark, concerns Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transients based on the PBMR-400 MW design. The deterministic neutronics, thermal-hydraulics and transient analysis tools and methods available to design and analyse PBMRs lag, in many cases, behind the state of the art compared to other reactor technologies. This has motivated the testing of existing methods for HTGRs but also the development of more accurate and efficient tools to analyse the neutronics and thermal-hydraulic behaviour for the design and safety evaluations of the PBMR. In addition to the development of new methods, this includes defining appropriate benchmarks to verify and validate the new methods in computer codes. The scope of the benchmark is to establish well-defined problems, based on a common given set of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark exercise has the following objectives: - Establish a standard benchmark for coupled codes (neutronics/thermal-hydraulics) for PBMR design; - Code-to-code comparison using a common cross section library ; - Obtain a detailed understanding of the events and the processes; - Benefit from different approaches, understanding limitations and approximations. Major Design and Operating Characteristics of the PBMR (PBMR Characteristic and Value): Installed thermal capacity: 400 MW(t); Installed electric capacity: 165 MW(e); Load following capability: 100-40-100%; Availability: ≥ 95%; Core configuration: Vertical with fixed centre graphite reflector; Fuel: TRISO ceramic coated U-235 in graphite spheres; Primary coolant: Helium; Primary coolant pressure: 9 MPa; Moderator: Graphite; Core outlet temperature: 900 C.; Core inlet temperature: 500 C.; Cycle type: Direct; Number of circuits: 1; Cycle

  5. Graphite selection for the PBMR reflector

    International Nuclear Information System (INIS)

    Marsden, B.J.; Preston, S.D.

    2000-01-01

    A high temperature, direct cycle gas turbine, graphite moderated, helium cooled, pebble-bed reactor (PBMR) is being designed and constructed in South Africa. One of the major components in the PBMR is the graphite reflector, which must be designed to last thirty-five full power years. Fast neutron irradiation changes the dimensions and material properties of reactor graphite, thus for design purposes a suitable graphite database is required. Data on the effect of irradiation on nuclear graphites has been gathered for many years, at considerable financial cost, but unfortunately these graphites are no longer available due to rationalization of the graphite industry and loss of key graphite coke supplies. However, it is possible, using un-irradiated graphite materials properties and knowledge of the particular graphite microstructure, to determine the probable irradiation behaviour. Three types of nuclear graphites are currently being considered for the PBMR reflector: an isostatically moulded, fine grained, high strength graphite and two extruded medium grained graphites of moderately high strength. Although there is some irradiation data available for these graphites, the data does not cover the temperature and dose range required for the PBMR. The available graphites have been examined to determine their microstructure and some of the key material properties are presented. (authors)

  6. Creation of the equilibrium core PBMR ORIGEN-S cross section library

    International Nuclear Information System (INIS)

    Stoker, C.C.; Reitsma, F.; Karriem, Z.

    2002-01-01

    As part of the design calculations for the Pebble Bed Modular Reactor (PBMR), fuel inventories, neutron and gamma sources and decay heat needs to be determined for the fuel spheres. Using the SCALE4.4 code system, a PBMR specific cross section library was created for the ORIGEN-S depletion calculations, assuming a 10-pass refueling system for the PBMR. In this paper the rationale for the creation of the PBMR library is evaluated in terms of the spectrum dependence due to burn-up. The ORIGEN-S PBMR library was further evaluated comparing the results for different parameters calculated with the reactor analysis diffusion code VSOP and the Monte Carlo code MCNP4C. (author)

  7. PBMR-SA licensing project organization

    International Nuclear Information System (INIS)

    Clapisson, G.A.; Metcalf, P.E.; Mysen, A.

    2001-01-01

    The South African nuclear regulatory authority, the Council for Nuclear Safety (CNS), is beginning the safety review of the Pebble Bed Modular Reactor (PBMR) design under development by the South African National Electrical Utility, Eskom. This paper describes the CNS licensing process, including the establishment of basic licensing criteria, general design criteria, and specific design rules, as well the safety assessment to be conducted in accordance with the established structure. It also summarises the CNS PBMR review project activities, including the overall organisational arrangements, licensing basis, safety and risk assessment, general operating rules and plant design engineering, and pre-operational testing. (author)

  8. The PBMR electric power generation plant; La planta de generacion de energia electrica PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Perez S, G.; Santacruz I, I.; Martin del Campo M, C. [FI-UNAM, 04500 Mexico D.F. (Mexico)] e-mail: gabriela_perez@engineer.com

    2003-07-01

    This work has as purpose to diffuse in a general way the technology of the one modulate reactor of pebble bed. Because our country is in developing ways, the electric power demand goes in increase with that which it is presented the great challenge of satisfying this necessity, not only being in charge of the one fact per se, but also involving the environmental aspect and of security. Both factors are covered by the PBMR technology, which we approach in their basic aspects with the purpose that the public opinion knows it and was familiarized with this type of reactors that well could represent a solution for our growing electricity demand. We will treat this reactor visualizing it like part of a generation plant defining in first place to the itself reactor. We will see because that the system PBMR consists of 2 main sections: the reactor and the unit of energy conversion, highlighting that the principle of the PBMR reactor operation is based on the thermodynamic Brayton cycle cooled by helium and that, in turn, it transmits the energy in form of heat toward a gas turbine. In what concerns to the fuel, it peculiar design due to its spherical geometry is described, aspect that make to this reactor different from the traditional ones that use fuel rods. In fact in the fuel spheres of the PBMR it is where it resides great part of it inherent security since each particle of fuel, consistent in uranium dioxide, is lined one with coal and silicon carbide those which form an impenetrable barrier containing to the fuel and those radioactive products that result of the nuclear reactions. Such particles are encapsulated in graphite to form the sphere or 'pebble', of here born the name of this innovative technology. (Author)

  9. Current status of the PBMR licensing project

    International Nuclear Information System (INIS)

    Mysen, A.; Clapisson, G.A.; Metcalf, P.E.

    2000-01-01

    The CNS is currently reviewing the PBMR conceptual design from a licensibility point of view. The PBMR concept is based on a High Temperature Gas Cooled Reactor - pebble bed reactor type. It is anticipated that the PBMR design will rely on inherent safety characteristics to contain fission products within fuel over the full range of design basis events. This feature combined with the high temperature integrity of the fuel and structural graphite, allows the safe use of a high coolant temperature, which allows consideration of the future development of this reactor for non-electrical applications of nuclear heat for industrial use. The CNS licensing approach requires that the licensing and design basis of the plant should respect prevailing international norms and practices and that a quantitative risk assessment should demonstrate compliance with the CNS fundamental safety standards. The first stage of the licensing process is now ongoing; this is a pre-application phase, which will result in a statement on licensibility being issued. Identification of the specific documentation requirements and information needed is required across every step of the licensing process. Top level regulatory requirements have been established for the PBMR. They include the CNS fundamental safety standard and basic licensing criteria, which describes requirements on licensees of nuclear installations regarding risk assessment and compliance with the safety criteria and define classification of licensing basis events. (author)

  10. The PBMR electric power generation plant

    International Nuclear Information System (INIS)

    Perez S, G.; Santacruz I, I.; Martin del Campo M, C.

    2003-01-01

    This work has as purpose to diffuse in a general way the technology of the one modulate reactor of pebble bed. Because our country is in developing ways, the electric power demand goes in increase with that which it is presented the great challenge of satisfying this necessity, not only being in charge of the one fact per se, but also involving the environmental aspect and of security. Both factors are covered by the PBMR technology, which we approach in their basic aspects with the purpose that the public opinion knows it and was familiarized with this type of reactors that well could represent a solution for our growing electricity demand. We will treat this reactor visualizing it like part of a generation plant defining in first place to the itself reactor. We will see because that the system PBMR consists of 2 main sections: the reactor and the unit of energy conversion, highlighting that the principle of the PBMR reactor operation is based on the thermodynamic Brayton cycle cooled by helium and that, in turn, it transmits the energy in form of heat toward a gas turbine. In what concerns to the fuel, it peculiar design due to its spherical geometry is described, aspect that make to this reactor different from the traditional ones that use fuel rods. In fact in the fuel spheres of the PBMR it is where it resides great part of it inherent security since each particle of fuel, consistent in uranium dioxide, is lined one with coal and silicon carbide those which form an impenetrable barrier containing to the fuel and those radioactive products that result of the nuclear reactions. Such particles are encapsulated in graphite to form the sphere or 'pebble', of here born the name of this innovative technology. (Author)

  11. Thermodynamic analysis of PBMR plant

    International Nuclear Information System (INIS)

    Sen, S.; Kadiroglu, O.K.

    2002-01-01

    The thermodynamic analysis of a PBMR is presented for various pressures and temperatures values. The design parameters of the components of the power plant are calculated and an optimum cycle for the maximum thermal efficiency is sought for. (author)

  12. Design of a physical model of the PBMR with the aid of Flownet

    International Nuclear Information System (INIS)

    Greyvenstein, G.P.; Rousseau, P.G.

    2002-01-01

    The design of a physical model of the PBMR with the aid of the code Flownet is discussed in this paper. The purpose of the physical model is to test the control strategies and operating procedures of the PBMR and also to demonstrate the accuracy of Flownet. Flownet is first used to do component matching and to determine the detail steady-state performance of the system. It is then demonstrated how the code was used to simulate the start-up procedure as well as a load following and a load rejection scenario. The study demonstrates how a micro model of the PBMR can be designed with the aid of a powerful simulation tool in a relatively short period of time and at low cost using commercially available turbochargers. (author)

  13. Xenon-induced axial power oscillations in the 400 MW PBMR

    International Nuclear Information System (INIS)

    Strydom, Gerhard

    2008-01-01

    The redistribution of the spatial xenon concentration in the 400 MW Pebble Bed Modular Reactor (PBMR) core has a non-linear, time-dependent feedback effect on the spatial power density during several types of operational transient events. Due to the inherent weak coupling that exists between the iodine and xenon formation and destruction rates, as well as the complicating effect of spatial variance in the thermal flux field, reactor cores have been analyzed for a number of decades for the occurrence and severity of xenon-induced axial power oscillations. Of specific importance is the degree of oscillation damping exhibited by the core during transients, which involves axial variations in the local power density. In this paper the TINTE reactor dynamics code is used to assess the stability of the current 400 MW PBMR core design with regard to axial xenon oscillations. The focus is mainly on the determination of the inherent xenon and power oscillation damping properties by utilizing a set of hypothetical control rod insertion transients at various power levels. The oscillation damping properties of two 100%-50%-100% load-follow transients, one of which includes the de-stabilizing axial effects of moving control rods, are also discussed in some detail. The study shows that, although first axial mode oscillations do occur in the 400 MW PBMR core, the inherent damping of these oscillations is high, and that none of the investigated load-follow transients resulted in diverging oscillations. It is also shown that the PBMR core exhibits no radial oscillation components for these xenon-induced axial power oscillations

  14. Steady-state and accident analyses of PBMR with the computer code SPECTRA

    International Nuclear Information System (INIS)

    Stempniewicz, Marek M.

    2002-01-01

    The SPECTRA code is an accident analysis code developed at NRG. It is designed for thermal-hydraulic analyses of nuclear or conventional power plants. The code is capable of analysing the whole power plant, including reactor vessel, primary system, various control and safety systems, containment and reactor building. The aim of the work presented in this paper was to prepare a preliminary thermal-hydraulic model of PBMR for SPECTRA, and perform steady state and accident analyses. In order to assess SPECTRA capability to model the PBMR reactors, a model of the INCOGEN system has been prepared first. Steady state and accident scenarios were analyzed for INCOGEN configuration. Results were compared to the results obtained earlier with INAS and OCTOPUS/PANTHERMIX. A good agreement was obtained. Results of accident analyses with PBMR model showed qualitatively good results. It is concluded that SPECTRA is a suitable tool for analyzing High Temperature Reactors, such as INCOGEN or for example PBMR (Pebble Bed Modular Reactor). Analyses of INCOGEN and PBMR systems showed that in all analyzed cases the fuel temperatures remained within the acceptable limits. Consequently there is no danger of release of radioactivity to the environment. It may be concluded that those are promising designs for future safe industrial reactors. (author)

  15. ANALISIS PENGENDALIAN DAYA REAKTOR PCMSR DENGAN LAJU ALIR PENDINGIN

    Directory of Open Access Journals (Sweden)

    Iqbal Syafin Noha

    2015-03-01

    Full Text Available Passive Compact Molten Salt Reactor (PCMSR merupakan pengembangan dari Molten Salt Reactor (MSR yang memiliki karakter berbeda dengan lima reaktor generasi IV lainnya, yaitu menggunakan bahan bakar leburan garam. Pada reaktor MSR, garam lebur tidak digunakan sebagai pendingin tetapi digunakan sebagai medium pembawa bahan bakar. Dengan fase bahan bakar yang berupa garam lebur LiF-BeF2-ThF4-UF4, maka dapat dilakukan pengendalian daya dengan mengatur laju aliran bahan bakar dan pendingin. Tujuan penelitian ini adalah untuk mengetahui pengaruh perubahan laju alir pendingin terhadap daya reaktor PCMSR. Analisis dilakukan dengan empat jenis masukan untuk perubahan laju alir pendingin, yaitu masukan step, ramp, eksponensial, dan sinusoidal. Untuk masukan step, laju alir pendingin dibuat berubah secara mendadak. Selanjutnya untuk masukan ramp dan eksponensal, perubahan laju alir masing-masing dibuat perlahan secara linear dan mengikuti fungsi eksponensial. Kemudian untuk masukan sinusoidal, laju alir berubah naik turun secara periodik dengan memvariasikan frekuensi dari perubahan laju alir tersebut. Hasil penelitian menunjukkan bahwa penurunan laju alir pendingin sebesar 50% dari laju pendingin sebelumnya, menyebabkan daya pada reaktor PCMSR turun sebesar 63% dari daya sebelumnya. Jika terjadi fluktuasi laju aliran pendingin, maka semakin cepat perubahan tersebut, maka respon daya yang diberikan semakin kecil. Pada frekuensi yang sangat cepat, daya reaktor menjadi konstan dan cenderung tidak memiliki respon terhadap laju aliran. Hal ini merupakan salah satu aspek keselamatan reaktor, karena reaktor tidak merespon perubahan yang terlalu cepat. Kemampuan reaktor mengatur daya menyesuaikan laju aliran pendingin merupakan aspek keselamatan lainnya. Kata kunci : PCMSR, pengendalian daya, laju alir pendingin, uji respon   Passive Compact Molten Salt Reactor (PCMSR is the development of Molten Salt Reactor (MSR which has different character from other five

  16. The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

    International Nuclear Information System (INIS)

    Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

    2006-01-01

    The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

  17. REAKTOR INNOVATIVE MOLTEN SALT (IMSR DENGAN SISTEM KESELAMATAN PASIF MENYELURUH

    Directory of Open Access Journals (Sweden)

    Andang Widiharto

    2015-04-01

    Full Text Available Pengembangan Teknologi Reaktor Nuklir pada masa mendatang mengarah pada peningkatan aspek keselamatan, peningkatan pendayagunaan bahan bakar, reduksi limbah radioaktif, ketahanan terhadap proliferasi bahan-bakar nuklir dan peningkatan aspek ekonomi. reaktor Innovative Molten Salt (IMSR adalah reaktor nuklir yang menggunakan bahan bakar cair berupa garam lebur fluoride (7LiF-ThF4-UF4-MaFx. Reaktor IMSR didesain sebagai reaktor pembiak termal, yaitu membiakkan U-233 dari Th-232. Hal ini untuk menjawab permasalahan sustainabilitas ketersedian sumber daya bahan bakar nuklir dan reduksi limbah radioaktif. Dalam aspek keselamatan, desain reaktor IMSR memiliki sifat inherent safe, yaitu koefisien umpan balik daya yang negatif serta memiliki fitur-fitur keselamatan pasif. Fitur-fitur keselamatan pasif terdiri dari sistem shutdown pasif, sistem pendinginan pasif pasca shutdown serta sistem pendinginan pasif untuk produk fisi. Kecelakaan yang berpotensi terjadi pada IMSR, yaitu kecelakaan kehilangan aliran bahan bakar, kecelakaan kehilangan aliran pendingin, kecelakaan kehilangan kemampuan pengambilan kalor serta kecelakaan kerusakan integritas sistem reaktor, dapat ditangani sepenuhnya secara pasif hingga mencapai kondisi shutdown selamat. Kata kunci: keselamatan pasif, inherent safe, IMSR   The next Nuclear Reactor Technology developments are directed to the increasing of the aspects of safety, fuel utility, radioactive waste reduction, proliferation retention and economy. Innovative Molten Salt Reactor (IMSR is a nuclear reactor design that uses fluoride molten salt (7LiF-ThF4-UF4-MaFx. IMSR is designed as a thermal breeder reactor, i.e. to produce U-233 from Th-232. This is the answer of natural nuclear fuel sustainability and radioactive waste problems. In term of safety aspect, IMSR design has inherent safe characteristics, i.e. negative power feedback coefficient, and passive safety features. The passive safety features are passive shutdown

  18. PENINGKATAN KINERJA SISTEM KESELAMATAN PASIF PADA REAKTOR NUKLIR DENGAN PENAMBAHAN KOMPONEN RVACS

    Directory of Open Access Journals (Sweden)

    A. G. Abdullah

    2014-07-01

    Full Text Available Kelengkapan sistem keselamatan pasif dan inheren pada reaktor lanjut merupakan prasyarat utama. Makalah ini mengeksplorasi hasil desain konseptual sistem pembuang sisa panas pada pusat listrik tenaga nuklir berjenis Very High-Temperature Reactor. Tujuan riset ini untuk merancang sistem pembuang sisa panas pusat listrik tenaga nuklir yang terdapat pada dinding reaktor. Studi kinerja Reactor Vessel Auxliary Cooling System (RVACS dilakukan pada dua jenis pendingin yaitu Timbal-Bismut dan Liquid Salt. Panas dari dinding reaktor dihapus melalui sirkulasi alamiah pada keadaan tunak. Analisis melibatkan sistem perpindahan panas secara radiasi, konduksi dan konveksi alami. Perhitungan perpindahan panas dilakukan pada elemen reaktor vessel, dinding luar guard vessel, dan pelat pemisah. Hasil analisis kecelakaan menunjukkan kedua jenis sistem pendingin reaktor dan sistem pasif sisa pembuangan panas cukup menghapus sisa panas hasil peluruhan dengan sirkulasi alami.ABSTRACTCompleteness of passive safety systems and inherent in advanced reactors is a major prerequisite. This paper explores the results of a conceptual design of the heat removal system at the nuclear power plant (NPP type Very High-Temperature Reactor. The purpose of this research was to design the reactor vessel auxiliary cooling system (RVACS of NPP located within the reactor walls. The RVACS performance study was conducted on two types of coolant: Lead-Bismuth and Liquid Salt. Heat was removed from the reactor vessel through the natural circulation in the steady state. Analyses of heat transfer systems involved radiation, conduction and natural convection. Heat transfer calculations were performed on the reactor vessel, guard vessel, and perforated plate. The results from the accident analysis showed that both types, the reactor coolant system and the passive residual heat removal system, adequately remove remaining heat of the decay by a natural circulation.

  19. Monte Carlo benchmark calculations for 400MWTH PBMR core

    International Nuclear Information System (INIS)

    Kim, H. C.; Kim, J. K.; Kim, S. Y.; Noh, J. M.

    2007-01-01

    A large interest in high-temperature gas-cooled reactors (HTGR) has been initiated in connection with hydrogen production in recent years. In this study, as a part of work for establishing Monte Carlo computation system for HTGR core analysis, some benchmark calculations for pebble-type HTGR were carried out using MCNP5 code. The core of the 400MW t h Pebble-bed Modular Reactor (PBMR) was selected as a benchmark model. Recently, the IAEA CRP5 neutronics and thermal-hydraulics benchmark problem was proposed for the testing of existing methods for HTGRs to analyze the neutronics and thermal-hydraulic behavior for the design and safety evaluations of the PBMR. This study deals with the neutronic benchmark problems, for fresh fuel and cold conditions (Case F-1), and first core loading with given number densities (Case F-2), proposed for PBMR. After the detailed MCNP modeling of the whole facility, benchmark calculations were performed. Spherical fuel region of a fuel pebble is divided into cubic lattice element in order to model a fuel pebble which contains, on average, 15000 CFPs (Coated Fuel Particles). Each element contains one CFP at its center. In this study, the side length of each cubic lattice element to have the same amount of fuel was calculated to be 0.1635 cm. The remaining volume of each lattice element was filled with graphite. All of different 5 concentric shells of CFP were modeled. The PBMR annular core consists of approximately 452000 pebbles in the benchmark problems. In Case F-1 where the core was filled with only fresh fuel pebble, a BCC(body-centered-cubic) lattice model was employed in order to achieve the random packing core with the packing fraction of 0.61. The BCC lattice was also employed with the size of the moderator pebble increased in a manner that reproduces the specified F/M ratio of 1:2 while preserving the packing fraction of 0.61 in Case F-2. The calculations were pursued with ENDF/B-VI cross-section library and used sab2002 S(α,

  20. Development and application of the PBMR fission product release calculation model

    International Nuclear Information System (INIS)

    Merwe, J.J. van der; Clifford, I.

    2008-01-01

    At PBMR, long-lived fission product release from spherical fuel spheres is calculated using the German legacy software product GETTER. GETTER is a good tool when performing calculations for fuel spheres under controlled operating conditions, including irradiation tests and post-irradiation heat-up experiments. It has proved itself as a versatile reactor analysis tool, but is rather cumbersome when used for accident and sensitivity analysis. Developments in depressurized loss of forced cooling (DLOFC) accident analysis using GETTER led to the creation of FIssion Product RElease under accident (X) conditions (FIPREX), and later FIPREX-GETTER. FIPREX-GETTER is designed as a wrapper around GETTER so that calculations can be carried out for large numbers of fuel spheres with design and operating parameters that can be stochastically varied. This allows full Monte Carlo sensitivity analyses to be performed for representative cores containing many fuel spheres. The development process and application of FIPREX-GETTER in reactor analysis at PBMR is explained and the requirements for future developments of the code are discussed. Results are presented for a sample PBMR core design under normal operating conditions as well as a suite of design-base accident events, illustrating the functionality of FIPREX-GETTER. Monte Carlo sensitivity analysis principles are explained and presented for each calculation type. The plan and current status of verification and validation (V and V) is described. This is an important and necessary process for all software and calculation model development at PBMR

  1. Reliability and integrity management program for PBMR helium pressure boundary components - HTR2008-58036

    International Nuclear Information System (INIS)

    Fleming, K. N.; Gamble, R.; Gosselin, S.; Fletcher, J.; Broom, N.

    2008-01-01

    The purpose of this paper is to present the results of a study to establish strategies for the reliability and integrity management (RIM) of passive metallic components for the PBMR. The RIM strategies investigated include design elements, leak detection and testing approaches, and non-destructive examinations. Specific combinations of strategies are determined to be necessary and sufficient to achieve target reliability goals for passive components. This study recommends a basis for the RIM program for the PBMR Demonstration Power Plant (DPP) and provides guidance for the development by the American Society of Mechanical Engineers (ASME) of RIM requirements for Modular High Temperature Gas-Cooled Reactors (MHRs). (authors)

  2. DESAIN KONSEPTUAL TERAS REAKTOR RISET INOVATIF BERBAHAN BAKAR URANIUM MOLIBDENUM DARI ASPEK NEUTRONIK

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2015-03-01

    Full Text Available Manfaat yang luas dari penggunaan reaktor riset membuat banyak negara membangun reaktor riset baru. Kecenderungan saat ini adalah reaktor tipe reaktor serbaguna (MPR dengan teras yang kompak untuk mendapatkan fluks neutron yang tinggi dengan daya yang relatif sedang atau rendah. Reaktor riset yang ada di Indonesia yang paling muda usianya sudah berumur 25 tahun. Oleh karena itu diperlukan desain reaktor riset baru sebagai alternatif, disebut reaktor riset inovatif (RRI, kelak pengganti reaktor riset yang sudah ada. Tujuan dari riset ini mendapatkan konfigurasi teras setimbang reaktor riset yang optimal dengan kriteria memiliki fluks neutron termal minimum sebesar 2,5x1014 n/cm2 s pada daya 20 MW (minimum, memiliki panjang operasi satu siklus lebih dari 40 hari dan penggunaan bahan bakar yang paling efisien. Desain neutronik dilakukan untuk bahan bakar baru U-9Mo-Al dengan kerapatan bervariasi dan jenis reflektor yang bervariasi. Desain dilakukan dengan paket program WIMSD-5B dan BATAN-FUEL. Hasil desain konseptual menyajikan 4 konfigurasi teras yaitu 5×5, 5×7, 6×5 dan 6×6. Hasil optimasi menunjukkan bahwa teras setimbang reaktor RRI dengan konfigurasi 5×5, tingkat muat 235U sebesar 450 g, reflektor berilium, fluks neutron termal maksimum di daerah reflektor sebesar 3,33×1014 neutron cm-2s-1 dan panjang siklus 57 hari merupakan desain teras reaktor riset inovatif yang paling optimal. Kata kunci: desain konseptual, bahan bakar uranium-molibdenum,berilium, D2O, WIMS, BATAN-FUEL   The multipurpose of research reactor utilization make many countries build the new research reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research newest reactor in Indonesia right now is already 25 year old. Therefore, it is needed to design a new research reactor, called innovative research reactor (IRR and then as an alternative to replace the old

  3. A study on the design concepts of the PBMR and the GT-MHR

    International Nuclear Information System (INIS)

    Park, Won Seok; Chang, Jong Hwa; Park, Chang Kue

    2004-05-01

    The major application of the nuclear power in the energy sector has been to produce the electricity. However, a growing concern on the environment and the expected shortage of the fossil energy resources is demanding the expansion of nuclear energy's role in the energy sectors. The High Temperature Gas cooled Reactor (HTGR) has been expected to expand the role of nuclear energy because of its high temperature capability. Especially, the interest on the HTGR has been sharply increased recently related with the production of the hydrogen. About 5 HTGRs had been operated by the end of 1980s. However, all of them had been terminated permanently at the end of 1980s because of their poor system economy and frequent technical troubles. A new concept called MHTGR (Modular High Temperature Gas cooled Reactor) emerged in early of 1990s. Two MHTGR concepts on commercial basis have been developed since then; one is the PBMR (Pebble Bed Modular Reactor) developed by Eskom in South Africa and another is GT-MHR (Gas Turbine Modular High-temperature Reactor) developed by both GA in USA and OKBM in Russia. In this report, the design concepts for the PBMR and GT-MHR were reviewed

  4. PBMR Project - Pebble Fuel Advantages

    International Nuclear Information System (INIS)

    Slabber, Johan; Matzie, Regis; Casperson, Sten; Kriel, Willem

    2006-01-01

    An overview is presented of all the important issues that influenced the choice of pebble fuel for the High-temperature Gas-cooled Reactor (HTGR) concept developed by South Africa. Each of these issues is then discussed in detail and compared with other fuel configurations proposed for direct cycle High-temperature Reactor (HTR) applications. The comparisons are provided using objective data generated by analyses done for the design of the Pebble Bed Modular Reactor (PBMR) and data that is available in open literature for the other fuel configurations

  5. ANALISIS PENGARUH DENSITAS BAHAN BAKAR SILISIDA TERHADAP PARAMETER KINETIK TERAS REAKTOR RSG-GAS

    Directory of Open Access Journals (Sweden)

    Tukiran s

    2016-11-01

    Full Text Available Saat ini RSG-GAS menggunakan elemen bakar silisida 2,96 g U/cc. Untuk meningkatkan waktu operasi reaktor maka akan direncanakan untuk mengganti elemen bakar silisida dengan kerapatan yang lebih tinggi. Keuntungan reaktor dengan bahan bakar kerapatan tinggi adalah dapat lebih efektif dan efisien. Maka perlu dilakukan perhitungan parameter kinetik teras silisida kerapatan tinggi mengingat pengaruhnya sangat penting untuk keselamatan operasi reaktor. Parameter kinetik yang dihitung yaitu fraksi neutron kasip efektif, konstanta peluruhan neutron kasip, umur neutron serempak yang merupakan faktor utama dalam kontrol dan keselamatan. Bahan bakar silisida tipe pelat dengan densitas 2,96 - 4,8 gU/cm3 digunakan pada teras RSG-GAS untuk menganalisis perhitungan parameter kinetik. Perhitungan sel dilakukan dengan paket program WIMSD-5B dan paket program Batan-2DIFF digunakan untuk perhitungan teras. Hasil perhitungan menunjukkan bahwa harga fraksi neutron kasip turun dengan naiknya densitas bahan bakar. Turunnya nilai parameter kinetik ini tidak mengganggu pergantian bahan bakar ke densitas yang lebih tinggi. Turunnya nilai parameter kinetik rata-rata dari densitas 2,96 gU/cm3 ke 3,55 gU/cm3 adalah 1,3 % sedangkan dari densitas 2,96 gU/cm3 ke 4,8 gU/cm3 adalah 2,2 % . Sehingga jika dilakukan pergantian bahan bakar maka ditinjau dari segi neutronik dan parameter kinetiknya tidak akan mengalami perubahan dalam pola operasi reaktor atau manajemen bahan bakar dan tidak akan berpengaruh terhadap keselamatan operasi reaktor.

  6. A study on the design concepts of the PBMR and the GT-MHR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Won Seok; Chang, Jong Hwa; Park, Chang Kue

    2004-05-01

    The major application of the nuclear power in the energy sector has been to produce the electricity. However, a growing concern on the environment and the expected shortage of the fossil energy resources is demanding the expansion of nuclear energy's role in the energy sectors. The High Temperature Gas cooled Reactor (HTGR) has been expected to expand the role of nuclear energy because of its high temperature capability. Especially, the interest on the HTGR has been sharply increased recently related with the production of the hydrogen. About 5 HTGRs had been operated by the end of 1980s. However, all of them had been terminated permanently at the end of 1980s because of their poor system economy and frequent technical troubles. A new concept called MHTGR (Modular High Temperature Gas cooled Reactor) emerged in early of 1990s. Two MHTGR concepts on commercial basis have been developed since then; one is the PBMR (Pebble Bed Modular Reactor) developed by Eskom in South Africa and another is GT-MHR (Gas Turbine Modular High-temperature Reactor) developed by both GA in USA and OKBM in Russia. In this report, the design concepts for the PBMR and GT-MHR were reviewed.

  7. Preliminary safety analysis of a PBMR supplying process heat to a co-located ethylene production plant

    International Nuclear Information System (INIS)

    Scarlat, Raluca O.; Cisneros, Anselmo T.; Koutchesfahani, Tawni; Hong, Rada; Peterson, Per F.

    2012-01-01

    This paper considers the safety analysis and licensing approach for co-locating a pebble bed modular reactor (PBMR) to provide process heat to an ethylene production unit. The PBMR is an advanced nuclear reactor design that provides 400 MW of thermal energy. Ethylene production is an energy intensive process that utilizes large gas furnaces to provide the heat for the process. Coupling a PBMR with an ethylene production plant would open a new market for nuclear power, and would provide the chemical industry with a cleaner power source, helping to achieve the Clean Air Act standards, and eliminating the 0.5 ton of CO 2 emissions per ton of produced ethylene. Our analysis uses the Chevron Phillips chemical plant in Sweeney, TX as a prototypical site. The plant has four ethylene production trains, with a total power consumption of 2.4 GW, for an ethylene output of 3.7 million tons per year, 4% of the global ethylene production capacity. This paper proposes replacement of the gas furnaces by low-emission PBMR modules, and presents the safety concerns and risk mitigation and management options for this coupled system. Two coupling design options are proposed, and the necessary changes to the design basis events and severe accidents for the PBMR licensing application are discussed. A joint effort between the chemical and the nuclear entities to optimize the coupling design, establish preventive maintenance procedures, and develop emergency response plans for both of the units is recommended.

  8. Preliminary safety analysis of a PBMR supplying process heat to a co-located ethylene production plant

    Energy Technology Data Exchange (ETDEWEB)

    Scarlat, Raluca O., E-mail: rscarlat@nuc.berkeley.edu [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States); Cisneros, Anselmo T. [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States); Koutchesfahani, Tawni [University of California, Chemical and Biomolecular Engineering, 201 Gilman Hall, Berkeley, CA 94720 (United States); Hong, Rada; Peterson, Per F. [University of California Berkeley, Nuclear Engineering, 4118 Etcheverry Hall, Berkeley, CA 94720 (United States)

    2012-10-15

    This paper considers the safety analysis and licensing approach for co-locating a pebble bed modular reactor (PBMR) to provide process heat to an ethylene production unit. The PBMR is an advanced nuclear reactor design that provides 400 MW of thermal energy. Ethylene production is an energy intensive process that utilizes large gas furnaces to provide the heat for the process. Coupling a PBMR with an ethylene production plant would open a new market for nuclear power, and would provide the chemical industry with a cleaner power source, helping to achieve the Clean Air Act standards, and eliminating the 0.5 ton of CO{sub 2} emissions per ton of produced ethylene. Our analysis uses the Chevron Phillips chemical plant in Sweeney, TX as a prototypical site. The plant has four ethylene production trains, with a total power consumption of 2.4 GW, for an ethylene output of 3.7 million tons per year, 4% of the global ethylene production capacity. This paper proposes replacement of the gas furnaces by low-emission PBMR modules, and presents the safety concerns and risk mitigation and management options for this coupled system. Two coupling design options are proposed, and the necessary changes to the design basis events and severe accidents for the PBMR licensing application are discussed. A joint effort between the chemical and the nuclear entities to optimize the coupling design, establish preventive maintenance procedures, and develop emergency response plans for both of the units is recommended.

  9. Optimization of inlet plenum of A PBMR using surrogate modeling

    International Nuclear Information System (INIS)

    Lee, Sang-Moon; Kim, Kwang-Yong

    2009-01-01

    The purpose of present work is to optimize the design of inlet plenum of PBMR type gas cooled nuclear reactor numerically using a combining of three-dimensional Reynolds-averaged Navier-Stokes (RANS) analysis and surrogate modeling technique. Shear stress transport (SST) turbulence model is used as a turbulence closure. Three geometric design variables are selected, namely, rising channel diameter to plenum height ratio, aspect ratio of the plenum cross section, and inlet port angle. The objective function is defined as a linear combination of uniformity of three-dimensional flow distribution term and pressure drop in the inlet plenum and rising channels of PBMR term with a weighting factor. Twenty design points are selected using Latin-hypercube method of design of experiment and objective function values are obtained at each design point using RANS solver. (author)

  10. ANALISIS POLA MANAJEMEN BAHAN BAKAR DESAIN TERAS REAKTOR RISET TIPE MTR

    Directory of Open Access Journals (Sweden)

    Lily Suparlina

    2015-03-01

    Full Text Available Parameter neutronik dibutuhkan dalam mendesain teras reaktor riset. Reaktor riset jenis MTR (Material Testing Reactor sangat diminati karena dapat digunakan baik untuk riset dan juga produksi radio isotop. Reaktor riset yang ada saat ini sudah tua sehingga dibutuhkan desain reaktor yang mempunyai teras kompak. Desain teras reaktor riset yang sudah ada saat ini belum cukup memadai untuk memenuhi persyaratan di dalam UCD yang telah ditetapkan yaitu fluks neutron termal di teras 1x1015 n/cm2s, oleh karena itu perlu dibuat desain teras reaktor baru sebagai alternatif yang kompak dan dapat menghasilkan fluks neutron tinggi. Telah dilakukan perhitungan dan analisis terhadap manajemen bahan bakar desain teras kompak dengan konfigurasi teras 5x5, berbahan bakar U9Mo-Al dan tinggi teras aktif 70 cm. Tujuan dari riset ini untuk memperoleh fluks neutron di teras memenuhi kebutuhan seperti yang telah ditetapkan di UCD dengan panjang siklus operasi minimum 20 hari pada daya 50 MW. Perhitungan dilakukan dengan menggunakan paket program komputer WIMSD-5B untuk menggenerasi tampang lintang makroskopik bahan bakar dan Batan-FUEL untuk memperoleh nilai parameter neutronik serta Batan-3DIFF untuk perhitungan nilai reaktivitas batang kendali. Perhitungan parameter neutronik teras reaktor riset ini dilakukan untuk bahan bakar U-9Mo-Al dengan tingkat muat bervariasi dan 2 macam pola pergantian bahan bakar yaitu teras segar dan teras setimbang. Hasil analisis menunjukkan bahwa pada teras segar, tingkat muat 235U sebesar 360 gram, 390 gram dan 450 gram memenuhi kriteria keselamatan dan kriteria penerimaan di UCD dengan nilai fluks neutron termal di teras lebih dari 1x1015 n/cm2s dan panjang siklus >20 hari, sedangkan pada teras setimbang panjang siklus dapat terpenuhi hanya untuk tingkat muat 450 gram. Kata kunci: desain teras reaktor, bahan bakar UMo, pola bahan bakar, WIMS, BATAN-FUEL   Research reactor core design needs neutronics parameter calculation use computer

  11. Introduction to the PBMR heat transfer test facility

    International Nuclear Information System (INIS)

    Rousseau, P.G.; Staden, M. van

    2008-01-01

    This paper provides an introduction to the Heat Transfer Test Facility (HTTF) that is currently being developed for PBMR (Pty.) Ltd. by M-Tech Industrial (Pty.) Ltd. in association with North-West University in South Africa. The paper provides an overview of the phenomena that will be studied and the envisaged test configurations for each of these phenomena. It also shows the layouts of the different test units namely the High Pressure Test Unit (HPTU) and the High Temperature Test Unit (HTTU) and provides an overview of the planned test schedule

  12. Test Results of PBMR Fuel Spheres

    International Nuclear Information System (INIS)

    Koshcheev, Konstantin; Diakov, Alexander; Beltyukov, Igor; Barybin, Andrey; Chernetsov, Mikhail

    2014-01-01

    Results of pre-irradiation testing of fuel spheres (FS) and coated particles (CP) manufactured by PBMR SOC (Republic of South Africa) are described. The stable high quality level of major characteristics (dimensions, CP coating structure, uranium-235 contamination of the FS matrix graphite and the outer PyC layer of the CP coating) are shown. Results of a methodical irradiation test of two FS in helium and neon medium at temperatures of 800 to 1300 °C with simultaneous determination of release-to-birth ratios for major gaseous fission products (GFP) are described. (author)

  13. ANALISIS DESAIN PROSES SISTEM PENDINGIN PADA REAKTOR RISET INOVATIF 50 MW

    Directory of Open Access Journals (Sweden)

    Sukmanto Dibyo

    2015-03-01

    Full Text Available Reaktor Riset Inovatif (RRI merupakan jenis MTR (Material Testing Reactor yang dipersiapkan ke depan sebagai desain reaktor baru. Daya RRI telah ditetapkan dari perhitungan neutronik dan termohidrolika teras yaitu 50 MW termal. Reaktor bertekanan 8 kgf/cm2 dan laju aliran massa pendingin primer 900 kg/s. Tantangan yang penting dalam menindak lanjuti desain reaktor ini adalah analisis desain pada sistem pendingin. Makalah ini bertujuan untuk menganalisis desain proses sistem pendingin utama reaktor RRI daya 50 MW (RRI-50 dengan menggunakan program Chemcad 6.1.4. Dalam analisis ini dilakukan perhitungan neraca massa dan energi (mass/energy balances pada sistem pendingin primer dan sekunder sebagai pendingin utama. Masing-masing sistem pendingin tersebut terdiri dari 2 jalur beroperasi secara paralel dan 1 jalur redundansi. Disamping itu untuk desain termal unit komponen telah dianalisis dengan program RELAP5, frenchcreek dan Metoda Analitik. Hasil analisis yang diperoleh adalah desain diagram sistem pendingin yang mencakup data parameter entalpi, temperatur, tekanan dan laju aliran massa pendingin untuk masing-masing jalur. Adapun hasil desain unit komponen utama pada RRI-50 adalah tangki tunda dengan volume 51,5 m3, 2 unit pompa sentrifugal dan 1 unit pompa cadangan pada pendingin primer daya 141 kW/pompa dan pendingin sekunder daya 206 kW/pompa, 2 unit penukar panas tipe shell-tube dengan koefisien termal overall 1377 W/m2.oC dan 4 unit menara pendingin yang mampu melepaskan panas ke udara dengan desain temperatur approach 5,0 oC dan temperatur range 9,0 oC. Desain sistem pendingin reaktor RRI-50 ini telah menetapkan parameter operasi sistem pendingin yaitu temperatur, tekanan dan laju aliran massa pendingin dengan mempertimbangkan tuntutan aspek keselamatan teras reaktor sehingga desain temperatur maksimum pendingin masuk ke teras 44,5 oC. Kata kunci : RRI 50 MW, desain sistem pendingin, program Chemcad 6.1.4   Innovative Research Reactor RRI

  14. Evolution of near term PBMR steam and cogeneration applications - HTR2008-58219

    International Nuclear Information System (INIS)

    Kuhr, R. W.; Hannink, R.; Paul, K.; Kriel, W.; Greyvenstein, R.; Young, R.

    2008-01-01

    US and international applications for large onsite cogeneration (steam and power) systems are emerging as a near term market for the PBMR. The South African PBMR demonstration project applies a high temperature (900 deg. C) Brayton cycle for high efficiency power generation. In addition, a number of new applications are being investigated using an intermediate temperature range (700-750 deg. C) with a simplified heat supply system design. This intermediate helium delivery temperature supports conventional steam Rankine cycle designs at higher efficiencies than obtained from water type reactor systems. These designs can be adapted for cogeneration of steam, similar to the design of gas turbine cogeneration plants that supply steam and power at many industrial sites. This temperature range allows use of conventional or readily qualifiable materials and equipment, avoiding some cost premiums associated with more difficult operating conditions. As gas prices and CO 2 values increase, the potential value of a small nuclear reactor with advanced safety characteristics increases dramatically. Because of its smaller scale, the 400-500 MWt PBMR offers the economic advantages of onsite thermal integration (steam, hot water and desalination co-production) and of providing onsite power at cost versus at retail industrial rates avoiding transmission and distribution costs. Advanced safety characteristics of the PBMR support the location of plants adjacent to steam users, district energy systems, desalination plants, and other large commercial and industrial facilities. Additional benefits include price stability, long term security of energy supply and substantial CO 2 reductions. Target markets include existing sites using gas fired boilers and cogeneration units, new projects such as refinery and petrochemical expansions, and coal-to-liquids projects where steam and power represent major burdens on fuel use and CO 2 emissions. Lead times associated with the nuclear licensing

  15. A CFD method to evaluate the integrated influence of leakage and bypass flows on the PBMR Reactor Unit

    International Nuclear Information System (INIS)

    Janse van Rensburg, J.J.; Kleingeld, M.

    2010-01-01

    Research highlights: → Research and analysis to identify and rank different leakage flow paths in a HTR. → Development of integrated CFD methodology for the prediction of leakage flows. → Development of a methodology to simulate flow resistances in above CFD model. → Validation of predicted flow results against different numerical methodology. → Illustration of the significant improvement achieved through this methodology. - Abstract: An area that has been identified as significantly important in the development of a High Temperature Reactor (HTR) is the prediction of leakage and bypass flows through such a reactor. It is therefore essential to understand the causes of bypass flows and to determine the effect on the predicted fuel and component temperatures. This paper discusses the identification of leakage flows that are applicable to the Pebble Bed Modular Reactor (Pty) Ltd. (PBMR) design and the ranking of these leakage flows. The modeling methodology and results are also discussed. Similar to previous HTR's, it was found that leakage and bypass flows are important parameters to consider for safe and efficient operation of the PBMR. Through a focused approach, it is shown that PBMR is able to improve the understanding of this phenomenon and quantify the flows and subsequent influence on the operation of the system. This has resulted in a reduction of leakage and bypass from approximately 46% to 20%. The improved understanding of leakage and bypass flows allows PBMR to address this issue during the design phase of the project, which subsequently results in a vast improvement over historical HTR designs. This gives PBMR a distinct advantage over previous High Temperature Reactors.

  16. Influence of temperature on the micro-and nano-structures of experimental PBMR TRISO coated particles: A comparative study - HTR2008-58189

    International Nuclear Information System (INIS)

    Van Rooyen, I. J.; Neethling, J. H.; Mahlangu, J.

    2008-01-01

    The PBMR fuel consists of TRISO Coated Particles (CPs) in a graphite matrix. The three layer system, IPyC-SIC-OPyC, forms the primary barrier to fission product release, with the SiC layer acting as the main pressure boundary of the particle. The containment of fission products inside the CPs is however a function of the operating temperature and microstructure of the SiC layer. During accident conditions, the CPs will reach higher temperatures than normal operating conditions. The Fuel Design Dept. of PBMR is therefore investigating various characteristics of the SiC layer, especially nano characteristics at variant conditions. The integrity of the interface between the SiC and Inner PyC layers is also important for fission product retention and therefore interesting TEM images of this region of the experimental PBMR TRISO particles will be shown. Transmission electron microscope (TEM) images of the microstructure of TRISO coated particles of three different experimental batches after annealing will be discussed. Particles annealed at 1980 deg. C for 1 hour revealed that the inner PyC layer de-bonded from the SiC layer. Changes observed in the diffraction rings are evidence that the PyC structure is becoming organized or anisotropic. The SiC layer, on the other hand, did not show any changes as a result of annealing. Only the cubic 3C-SiC phase was observed for a limited number of grains analyzed. The nano hardness and elasticity measurements of the three test batches were done using a CSM Nano Hardness Tester. These results are compared to indicate possible differences between the 1 hour and 5 hour annealing time as a function of annealing temperature from 1000 deg. C to 1980 deg. C. The variation of hardness and elasticity as a function of temperature for the three experimental batches are identified and discussed. This preliminary TEM investigation and nano hardness measurements have contributed new knowledge about the effect of high temperature annealing on

  17. Licensing of the proposed PBMR-SA

    International Nuclear Information System (INIS)

    Clapisson, G.A.; Henderson, N.R.; Hill, T.F.; Keenan, N.H.; Metcalf, P.E.; Mysenkov, A.

    1997-01-01

    This paper describes the preliminary criteria, which are intended to be used by the South African regulatory authority (CNS), for licensing of the ESKOM proposed - South African high temperature gas-cooled Pebble Bed Modular Reactor (PBMR-SA). The CNS intends to apply the existing CNS licensing approach together with some international design criteria used for advanced reactors as well as international experience gained from the safety evaluation of the MHTGR, THTR, etc. A major requirement to this type of reactor is that it should comply with the current CNS risk criteria and provide, as a minimum, the same degree of protection to the operator, public and environment that is required for the current generation of nuclear reactors. (author)

  18. UJI COBA TEKNOLOGI BIOFILM KONSORSIUM BAKTERI PADA REAKTOR SEMIANAEROB-AEROB UNTUK PENGOLAHAN AIR LIMBAH DI INDUSTRI PENCELUPAN TEKSTIL SKALA RUMAH TANGGA

    Directory of Open Access Journals (Sweden)

    Dewa Ketut Sastrawidana

    2013-04-01

    Full Text Available Penelitian ini bertujuan untuk menganalisis efektifitas teknologi biofilm konsorsium bakteri pada  reaktor semianaerob-aerob ntuk mengolah air limbah pencelupan tekstil. Bakteri pada reaktor semianaerob terdiri dari  Aeromonas sp. Pseudomonas sp, dan Flavobacterium sp. sedangkan pada reaktor aerob terdiri dari Vibrio sp. Plesiomonas sp. dan Enterobacter sp. Perombakan proses pertumbuhan terlekat diawali dengan menumbuhkan konsorsium bakteri pada masing-masing reaktor selama 10  hari menggunakan pada batu vulkanik merah sebagai media pelekatan bakteri. Setelah terbentuk biofilm,selanjutnya digunakan untuk merombak limbah denagn waktu tinggal limbah 2 hari. Hasil penelitian menunjukkan teknologi biofilm cukup efektif diaplikasikan pada skala lapang menghasilkan efisiensi perombakan TSS, BOD dan COD secara berturut-turut sebesar 84,7%; 80,56% dan 90,40%. Uji toksisitas air limbah tekstil menggunakan ikan nila dengan waktu paparan 3 hari menunjukkan bahwa air limbah tekstil sebelum diolah berkatagori toksik ringan dengan nilai EC50 adalah 88,80% sedangkan setelah diolah dalam reaktor biofilm konsorsium bakteri sistem anaerob-aerob selama 2 hari menjadi katagori tidak toksik dengan nilai EC50 sebesar 101,64%. Dengan demikian, pengolahan limbah tektil dengan sistem kombinasi anaerob-aerob menghasilkan kualitas limbah dengan kriteria sudah memenuhi baku mutu untuk dibuang ke lingkungan.

  19. ANALISIS SEBARAN RADIONUKLIDA PADA KONDISI NORMAL UNTUK REAKTOR AEC 1000 MW

    Directory of Open Access Journals (Sweden)

    Sri Kuntjoro

    2015-03-01

    Full Text Available Telah dilakukan analisis sebaran radionuklida pada reaktor daya Atomic Energy Agency (AEC 3568 MWTh, setara dengan 1000 Mwe untuk kondisi operasi normal. Analisis dilakukan untuk dua reaktor yang terpisah sejauh 500 m dan sudut 90o satu dengan yang lain. Langkah awal dalam melakukan analisis adalah menentukan suku sumber reaktor menggunakan program komputer ORIGEN2 dan EMERALD NORMAL. ORIGEN2 digunakan untuk menentukan inventori radionuklida yang terdapat di reaktor. Selanjutnya dengan dengan menggunakan program EMERALD NORMAL dihitung suku sumber yang sampai ke cerobong reaktor. Untuk menganalisis dosis yang diterima penduduk dilakukan dengan menggunakan program PC-CREAM. Perhitungan dilakukan untuk satu dan dua PLTN di calon tapak PLTN. Hasil yang diperoleh adalah sebaran radionuklida terbesar untuk satu PLTN pada jarak 1 km dan kearah zona 9 (191,25o dan untuk dua PLTN pada jarak 1 km dan kearah zona 10 (213,75o. Radionuklida yang sampai ke penduduk melalui dua alur yaitu alur makanan dan hirupan. Untuk alur makanan berasal dari radionuklida I-131, dan terbesar melalui alur produk susu sebesar 53,40 % untuk satu maupun dua PLTN . Untuk alur hirupan ranionuklida pemberi kontribusi paparan terbesar berasal dari Kr-85m sebesar 53,80 %. Dosis total terbesar yang diterima penduduk terdapat pada jarak 1 Km untuk bayi yaitu sebesar 4,10 μSi dan 11,26 μSi untuk satu dan dua PLTN. Hasil ini sangat kecil dibandingkan dengan batas dosis yang diijinkan oleh badan pengawas (BAPETEN untuk penduduk yaitu sebesar 1 mSi. Kata Kunci : Reaktor daya, komputer code, radionuklida, alur makanan, hirupan   Analysis for radionuclide dispersion for the Atomic Energy Agency (AEC 3568 MWth Power Reactor, equal to the 1000 MWe at normal condition has been done. Analysis was done for two piles that is separated by 500 m distance and angle of 90o one to other. Initial pace in doing the analysis is to determine reactors source term using ORIGEN2 and EMERALD NORMAL

  20. Design and simulation of a process of seawater desalination (MED) using the residual heat of a PBMR nuclear power plant; Diseno y simulacion de un proceso de desalinizacion de agua de mar (MED) utilizando el calor residual de una planta nucleoelectrica PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, Julio; Morales S, J.B. [UNAM, DEPFI Campus Morelos, Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

    2008-07-01

    In the present work it is demonstrated as the thermodynamic recuperative Brayton cycle with which operates a nuclear power plant type PBMR (Pebble Bed Modular Reactor) it allows to use the residual heat, removed in the coolers of the compression stage of the system, to produce vapor and to desalt seawater. The desalination process selected, starting from its operation characteristics and the derived advantages of them using nuclear heat, it the one of Multi-Effect Distillation, MED for its abbreviations in English, which described and it is justified to detail. This distillation process widely studied, allows us to use water vapor pressurized to temperatures between 70 and 110 C like energy source to evaporate the seawater in the first stage or effect of the process. The relatively low temperatures with which the vapor takes place of feeding to the process is it makes to the plant PBMR ideal for desalination of seawater, since does not require majors modifications to its design its operation, and on the contrary it allows to use the heat that previously was rejected, to produce the vapor. In this work an unit MED of six effects is designed, which undergoes a successive fall of pressure in each of them. Once obtained the agreed design to the conditions of operation of PBMR plant, it was model mathematically the MED process, including the coupling stage with the reactor coolers. The mathematical model was obtained by means of differential equations of mass balance and energy in the system, and with this it was implemented in SIMULINK a model equivalent to the MED process which is interconnected to the simulator coolers of the PBMR plant, constructed previously. One ran the program being obtained the results that are reported at the end of this article. (Author)

  1. Effect of high temperature annealing on the grain size of CVD-grown SiC and experimental PBMR TRISO coated particles

    CSIR Research Space (South Africa)

    Mokoduwe, SM

    2010-10-01

    Full Text Available in the PBMR fuel SiC layer. square samples were cut from the original sample received from ORNL and prepared for grain size Prague, Czech Republic, October 18 – 2000 °C. These no significant ion of how the 8] also ge is also of tal THODS -Si... for grain size determination Fig. 5: Influence of high temperature annealing on the CVD ORNL polycrystalline 3 C-SiC. Fig. 6: Influence of high temperature annealing on the polycrystalline 3 C-SiC layer of PBMR TRISO CP batches D and E...

  2. ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-06-01

    Full Text Available ABSTRAK ANALISIS KONDISI TERAS REAKTOR DAYA MAJU AP1000 PADA KECELAKAAN SMALL BREAK LOCA. Kecelakaan yang diakibatkan oleh kehilangan pendingin (loss of coolant accident / LOCA dari sistem reaktor merupakan kejadian dasar desain yang tetap diantisipasi dalam desain reaktor daya yang mengadopsi teknologi Generasi II hingga IV. LOCA ukuran kecil (small break LOCA memiliki dampak yang lebih signifikan terhadap keselamatan dibandingkan LOCA ukuran besar (large break LOCA seperti terlihat pada kejadian Three-Mile Island (TMI. Fokus makalah adalah pada analisis small break LOCA pada reaktor daya maju Generasi III+ yaitu AP1000 dengan mensimulasikan tiga kejadian pemicu yaitu membukanya katup Automatic Depressurization System (ADS secara tak disengaja, putusnya salah satu pipa Direct Vessel Injection (DVI secara double-ended, dan putusnya pipa lengan dingin dengan diameter bocoran 10 inci. Metode yang digunakan adalah simulasi kejadian pada model AP1000 yang dikembangkan secara mandiri menggunakan program perhitungan RELAP5/SCDAP/Mod3.4. Dampak yang ingin dilihat adalah kondisi teras selama terjadinya small break LOCA yang terdiri dari pembentukan mixture level dan transien temperatur kelongsong bahan bakar. Hasil simulasi menunjukkan bahwa mixture level untuk semua kejadian small break LOCA berada di atas tinggi teras aktif yang menunjukkan tidak terjadinya core uncovery. Adanya mixture level berpengaruh pada transien temperatur kelongsong yang menurun dan menunjukkan pendinginan bahan bakar yang efektif. Hasil di atas juga identik dengan hasil perhitungan program lain yaitu NOTRUMP. Keefektifan pendinginan teras juga disebabkan oleh berfungsinya injeksi pendingin melalui fitur keselamatan pasif yang menjadi ciri reaktor daya AP1000. Secara keseluruhan, hasil analisis menunjukkan model AP1000 yang telah dikembangkan dengan RELAP5 dapat digunakan untuk keperluan analisis kecelakaan dasar desain pada reaktor daya maju AP1000. Kata kunci: analisis

  3. PEMODELAN TERAS UNTUK ANALISIS PERHITUNGAN KONSTANTA MULTIPLIKASI REAKTOR HTR-PROTEUS

    Directory of Open Access Journals (Sweden)

    Zuhair Zuhair

    2015-04-01

    Full Text Available PTRKN sebagai salah satu unit kerja di BATAN dengan tugas pokok dan fungsi yang berkaitan erat dengan teknologi reaktor dan keselamatan nuklir, menaruh perhatian khusus pada konsep reaktor pebble bed. Dalam makalah ini pemodelan reaktor pebble bed HTR-PROTEUS dilakukan dengan program transport Monte Carlo MCNP5. Partikel bahan bakar berlapis TRISO dimodelkan secara detail dan eksak dimana distribusi acak partikel ini dalam bola bahan bakar didekati menggunakan array teratur kisi SC dengan fraksi packing 5,76% tanpa zona eksklusif. Model teras pebble bed didekati dengan memanfaatkan kisi teratur dari bola yang disusun sebagai kisi BCC berdasarkan sel berulang yang digenerasi dari sejumlah sel satuan. Hasil perhitungan MCNP5 memperlihatkan kesesuaian yang sangat baik dengan eksperimen, walaupun teras HTR-PROTEUS diprediksi lebih reaktif daripada pengukuran, khususnya di teras 4.2 dan 4.3. Pustaka ENDF/B-VI menunjukkan konsistensi dengan estimasi keff paling akurat dibandingkan pustaka ENDF/B-V, terutama ENDF/B-VI (66c. Deviasi estimasi keff yang dihitung dengan eksperimen dikaitkan sebagai konsekuensi dari komposisi reflektor grafit yang dispesifikasikan. Komparasi yang dibuat memperlihatkan bahwa MCNP5 menghasilkan keff teras HTR-PROTEUS lebih presisi daripada hasil dari MCNP4B dan MCNPBALL. Hasil ini menyimpulkan bahwa, sukses metodologi pemodelan ini menjustifikasi aplikasi MCNP5 untuk analisis reaktor pebble bed lainnya. Kata kunci: pemodelan teras HTR-PROTEUS, konstanta multiplikasi, MCNP5   PTRKN as a working unit in BATAN whose main duties and functions are related to reactor technology and nuclear safety, consern attention to pebble bed reactor concept. In this paper modeling of HTR-PROTEUS pebble bed reactor was done using Monte Carlo transport code MCNP5. The TRISO coated fuel particle is modeled in detailed and exact manner where random distributions of these particles in fuel pebble is approximated by using regular array of SC lattice

  4. Risk, probability and uncertainty in the calculations of gas cooled reactor of PBMR type. Part 2

    International Nuclear Information System (INIS)

    Serbanescu, Dan

    2004-01-01

    The paper presents the main conclusions of the insights to a cooled gas reactor from the perspective of the following notions: probability, uncertainty, entropy and risk. Some results of the on-going comparison between the insights obtained from three models and approaches are presented. The approaches consider the Pebble Bed Module Reactor (PBMR) NPP as a thermodynamic installation and as hierarchical system with or without considering the information exchange between its various levels. The existing model was a basis for a PRA going on in phases for PBMR. In the first part of this paper results from phase II of this PRA were presented. Further activities going on in the preparation for phase II PRA and for the development of a specific application of using PRA during the design phases for PBMR are undergoing with some preliminary results and conclusions. However, for the purposes of this paper and the comparative review of various models in the part two one presents the risk model (model B) based on the assumption and ideas laid down at the basis of the future inter-comparison of this model with other plant models. The assumptions concern: the uncertainties for the quantification of frequencies; list of initiated events; interfaces with the deterministic calculation; integrated evaluation of all the plant states; risk of the release of radionuclide; the balance between the number and function of the active systems and the passive systems; systems interdependencies in PBMR PRA; use of PRA for the evaluation of the impact of various design changes on plant risk. The model B allows basically evaluating the level of risk of the plant by calculating it as a result of acceptance challenge to the plant. By using this model the departure from a reference state is given by the variation in the risk metrics adopted for the study. The paper present also the synergetic model (model C). The evaluation of risk in the model C is considering also the information process. The

  5. Support to design and construction of the PBMR plant

    International Nuclear Information System (INIS)

    Cazorla, F.; Moron, P.; Gonzalez, J. I.

    2010-01-01

    Developing the new reactor design to a licensable state for constructing a pilot plant is a tough task require specific resources, concerning knowledge and previous experience, which trespassing the pure scientific or technologic knowledge linked to the reactor conceptual design. Taking into consideration the experience derived from the collaboration between the South African company PBMR (PTY) Ltd.; the Pebble Bed Modular Reactor Designer, and Tecnatom SA, the article presents some of the aspects in which the companies or organization in charge of the design can demand external support to license and construct the pilot plants with guaranteed success. (Author)

  6. The pebble bed modular reactor (PBMR) as a source of high quality process heat for sustainable oil sands expansion

    International Nuclear Information System (INIS)

    Morris, A.; Kuhr, R.

    2008-01-01

    Bitumen extraction, processing and upgrading consumes large quantities of natural gas for production of steam, hot water and hydrogen. Massive expansion of bitumen production is planned in response to energy demands, oil prices, and the desire for energy security. The PBMR in its Process Heat configuration supports applications that compete in a cost effective and environmentally sustainable way with natural gas fired boilers and steam methane reforming. The PBMR has the benefit of size, passive nuclear safety characteristics (encompassing Generation IV safety principles), high reliability, high temperature process heat (750-950 o C) in a modular design suited to the oil sands industry. (author)

  7. PBMR spent fuel bulk dry storage heat removal - HTR2008-58170

    International Nuclear Information System (INIS)

    De Wet, G. J.; Dent, C.

    2008-01-01

    A low decay heat (implying Spent Fuel (SF) pebbles older than 8-9 years) bulk dry storage section is proposed to supplement a 12-tank wet storage section. Decay heat removal by passive means must be guaranteed, taking into account the fact that dry storage vessels are under ground and inside the building footprint. Cooling takes place when ambient air (drawn downwards from ground level) passes on the outside of the 6 tanks' vessel containment (and gamma shielding), which is in a separate room inside the building, but outside PBMR building confinement and open to atmosphere. Access for loading/unloading of SF pebbles is only from the top of a tank, which is inside PBMR building confinement. No radioactive substances can therefore leak into atmosphere, as vessel design will take into account corrosion allowance. In this paper, it is shown (using CFD (Computational Fluid Dynamics) modelling and analytical analyses) that natural convection and draught induced flow combine to remove decay heat in a self-sustaining process. Decay heat is the energy source, which powers the draught inducing capability of the dry storage modular cell system: the more decay heat, the bigger the drive to expel heated air through a higher outlet and entrain cool ambient air from ground level to the bottom of the modular cell. (authors)

  8. ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI

    Directory of Open Access Journals (Sweden)

    Azizul Khakim

    2015-10-01

    Full Text Available ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET.   ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the

  9. A multi-tank storage facility to effect power control in the PBMR power cycle

    International Nuclear Information System (INIS)

    Matimba, T.A.D.; Krueger, D.L.W.; Mathews, E.H.

    2007-01-01

    This article presents the concept of a storage facility used to effect power control in South Africa's PBMR power cycle. The concept features a multiple number of storage vessels whose purpose is to contain the working medium, helium, as it is withdrawn from the PBMR's closed loop power cycle, at low energy demand. This helium is appropriately replenished to the power cycle as the energy demand increases. Helium mass transfer between the power cycle and the storage facility, henceforth known as the inventory control system (ICS), is carried out by way of the pressure differential that exists between these two systems. In presenting the ICS concept, emphasis is placed on storage effectiveness; hence the discussion in this paper is centred on those features which accentuate storage effectiveness, namely:- Storage vessel multiplicity; - Unique initial pressures for each vessel arranged in a cascaded manner; and - A heat sink placed in each vessel to provide thermal inertia. Having presented the concept, the objective is to qualitatively justify the presence of each of the above-mentioned features using thermodynamics as a basis

  10. Analysis of PBMR transients using a coupled neutron transport/thermal-hydraulics code DORT-TD/thermix

    International Nuclear Information System (INIS)

    Tyobeka, B.; Ivanov, K.; Pautz, A.

    2007-01-01

    In the advent of increased demand for safety and economics of nuclear power plants, nuclear engineers and designers are called upon to develop advanced computation tools. In these developments, space-time effects in the dynamics of nuclear reactors must be considered within the framework of a full 3-dimensional treatment of both neutron kinetics and thermal hydraulics. In a recent effort at the Pennsylvania State University, a time-dependent version of the discrete ordinates transport code DORT, DORT-TD was coupled to a 2-dimensional core thermal hydraulics code THERMIX-DIREKT. In the coupling process, a feedback model was developed to account for the feedback effects and was implemented into DORT-TD. During the calculation process for each spatial node of the DORT-TD core model, feedback parameters representative of this node are passed to the feedback module. Using these values, cross section tables are then interpolated for the appropriate macroscopic cross section values. The updated macroscopic cross sections are passed back to DORT-TD to perform transport core calculations, and the power distribution is transferred to THERMIX-DIREKT to obtain the relevant thermal-hydraulics data in turn, and this calculation loop continues. In this paper, DORT-TD/THERMIX is used to simulate transients of interest in the PBMR (Pebble Bed Modular Reactor) safety using established benchmark problems: load change from 100% to 40% power and fast control rod ejection (PBMR-268 benchmark problem). The results obtained are compared with those obtained using the diffusion-based module of the code. The results are only preliminary and so far show that diffusion theory is not such a bad approximation for PBMR for the prediction of integral parameters

  11. The OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark - Steady-state results and status

    International Nuclear Information System (INIS)

    Reitsma, F.; Han, J.; Ivanov, K.; Sartori, E.

    2008-01-01

    The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated. Since only a few pebble-bed HTR experimental facilities or plant data are available the use of code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MW design and a representative set of transient cases is defined as an OECD benchmark. The scope of the benchmark is to establish a series of well-defined multi-dimensional computational benchmark problems with a common given set of cross-sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. The OECD benchmark includes steady-state and transients cases. Although the focus of the benchmark is on the modelling of the transient behaviour of the PBMR core, it was also necessary to define some steady-state cases to ensure consistency between the different approaches before results of transient cases could be compared. This paper describes the status of the benchmark project and shows the results for the three steady state exercises defined as a standalone neutronics calculation, a standalone thermal-hydraulic core calculation, and a coupled neutronics/thermal-hydraulic simulation. (authors)

  12. Verification of Serpent code for the fuel analysis of a PBMR; Verificacion del codigo SERPENT para el analisis de combustible para un PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper the models and simulations with the Monte Carlo code Serpent are presented, as well as the obtained results of the different analyzed cases in order to verify the suitability or reliability of the use of this code to ensure favorable results in the realization of a neutronic analysis of fuel for a Pebble Bed Modular Reactor (PBMR). Comparisons were made with the results reported in a report by the OECD/Nea relative to a high temperature reactor of spheres bed with plutonium reactor grade as fuel. The results show that the use of Serpent is appropriate, as these results are comparable with those reported in the report. (Author)

  13. Challenges and opportunities in providing a digital protection system for the PBMR - HTR2008-58173

    International Nuclear Information System (INIS)

    Marais, J.; Ridolfo, C. F.

    2008-01-01

    The Republic of South Africa is currently developing the Pebble Bed Modular Reactor (PBMR); an advanced, fourth-generation reactor that incorporates inherent safety features, which require no human intervention and which provide an unprecedented level of nuclear safety. In addition to electrical power generation, the reactor is uniquely suited for a variety of non-traditional nuclear applications including oil sands extraction, desalination, and hydrogen production. A state-of-the-art digital Protection System for the PBMR is currently being developed in conjunction with Westinghouse Electric Company (WEC). The Protection System provides for: reactor shutdown using two different reactor trip methodologies (dropping of the control rods and insertion of Small Absorber Spheres (SASs) which are composed of boron carbide); post-event monitoring; and manual reactor shutdown, which is independent of software-based systems. The reactor shutdown and post-event instrumentation monitoring components of the Protection System are being implemented utilizing the WEC 'Common Q' platform, which is comprised of 'commercially dedicated' Programmable Logic Controllers (PLCs), colour-graphic Flat Panel Displays (FPDs) with integral touch screens, and high-speed data communication links. High reliability and availability are achieved through component redundancy, continuous automatic self-testing which is run online in a background mode, and implementation of a multi-channel system design which is tolerant to failures. The Protection System is also designed to support periodic surveillance testing through a suite of built- in computer-aided test facilities that are accessible via an FPD interface. These allow various system surveillance requirements to be readily performed in a convenient and systematic manner. This paper discusses the following topics with regard to the PBMR Protection System: development strategy, functional requirements, selection of applicable Codes and Standards

  14. The first stage of licensing of PBMR in South Africa and safety issues

    International Nuclear Information System (INIS)

    Clapisson, G.A.; Mysen, A.

    2002-01-01

    The National Nuclear Regulator (NNR) has received a nuclear installation licence application from Eskom (the South African electricity utility). The Application is made in accordance with the National Nuclear Regulator Act for a nuclear installation licence for the demonstration module of a 110 MWe Class Pebble Bed Modular Reactor (PBMR) electricity generating power station. It is proposed to locate the installation on Eskom property within the owner-controlled boundary of Koeberg Nuclear Power Station situated in the Western Cape, subject to inter alia a favourable Environmental Impact Assessment (EIA) record of decision, which is currently being undertaken under the requirements of another legislation the Environment Conservation Act. The PBMR is a graphite moderated helium cooled reactor using a direct gas cycle to convert the heat, generated by nuclear fission in the reactor and transferred to the coolant gas, into electrical energy by means of a helium turbo-generator. By design, provision has been made to accommodate the storage of spent fuel in the buildings for the 40-year design life of the plant and thereafter for a further period if so required. Radioactive material and waste will be managed and disposed of in accordance with Regulatory and Government legal requirements. (authors)

  15. The OECD/NEA/NSC PBMR 400 MW coupled neutronics thermal hydraulics transient benchmark: transient results - 290

    International Nuclear Information System (INIS)

    Strydom, G.; Reitsma, F.; Ngeleka, P.T.; Ivanov, K.N.

    2010-01-01

    The PBMR is a High-Temperature Gas-cooled Reactor (HTGR) concept developed to be built in South Africa. The analysis tools used for core neutronic design and core safety analysis need to be verified and validated, and code-to-code comparisons are an essential part of the V and V plans. As part of this plan the PBMR 400 MWth design and a representative set of transient exercises are defined as an OECD benchmark. The scope of the benchmark is to establish a series of well defined multi-dimensional computational benchmark problems with a common given set of cross sections, to compare methods and tools in coupled neutronics and thermal hydraulics analysis with a specific focus on transient events. This paper describes the current status of the benchmark project and shows the results for the six transient exercises, consisting of three Loss of Cooling Accidents, two Control Rod Withdrawal transients, a power load-follow transient, and a Helium over-cooling Accident. The participants' results are compared using a statistical method and possible areas of future code improvement are identified. (authors)

  16. Fire modeling of the Heiss Dampf Reaktor containment

    International Nuclear Information System (INIS)

    Nicolette, V.F.; Yang, K.T.

    1995-09-01

    This report summarizes Sandia National Laboratories' participation in the fire modeling activities for the German Heiss Dampf Reaktor (HDR) containment building, under the sponsorship of the United States Nuclear Regulatory Commission. The purpose of this report is twofold: (1) to summarize Sandia's participation in the HDR fire modeling efforts and (2) to summarize the results of the international fire modeling community involved in modeling the HDR fire tests. Additional comments, on the state of fire modeling and trends in the international fire modeling community are also included. It is noted that, although the trend internationally in fire modeling is toward the development of the more complex fire field models, each type of fire model has something to contribute to the understanding of fires in nuclear power plants

  17. Analysis of Dust and Fission Products in PBMR Turbine

    International Nuclear Information System (INIS)

    Stempniewicz, M.M.; Wessels, D.

    2014-01-01

    A 400 MWth direct cycle Pebble Bed Modular reactor was under development in South Africa. The work performed included design and safety analyses. In HTR/PBMR, graphite dust is generated during normal reactor operation due to pebble-to-pebble scratching. This dust will be deposited throughout the primary system. Furthermore, the dust will become radioactive due to sorption of fission products released, although in very small quantities, during normal operation. This paper presents a model and analyses of the PBMR turbine with the SPECTRA code. The purpose of the present work was to estimate the amount and distribution of deposited dust and the fission products, namely cesium, iodine, and silver, during plant life-time, which was assumed to be 40 full-power years. The performed work showed that after 40 years of plant life-time deposited layers are very small. The largest deposition is of course observed on the dust filters. Apart from the dust filters, the largest dust deposition is observed on the: • Outer Casing (inner walls) • Turbine Rotor Cooling Cavity (inner walls) • HPC Cold Cooling Gas Header (inner walls) This is caused by relatively low gas velocities in these volumes. The low velocities allow a continuous build-up of the dust layer. About 90% of cesium, 40% of iodine, and 99.9% of silver is adsorbed on the metallic structures of the turbine. The sorption rate increases along the turbine due to decreasing temperatures. In case of cesium and iodine the highest concentrations are observed in the last stage (stage 12) of the turbine. In the case of silver the sorption is so large that the silver vapor is significantly depleted in the last stages of the turbine. This is a reason for having a maximum in silver concentration in the stage 10. In the following stages the concentration decreases due to very small silver vapor fraction in the gas. (author)

  18. Neutronic analysis of the PBMR-400 full core using thorium fuel mixed with plutonium or minor actinides

    International Nuclear Information System (INIS)

    Acır, Adem; Coşkun, Hasan

    2012-01-01

    Highlights: ► Neutronic calculations for PBMR 400 were conducted with the computer codes MCNP and MONTEBURNS 2.0. ► The criticality and burnup were investigated for reactor grade plutonium and minor actinides. ► We found that the use of these new fuels in PBMRs would reduce the nuclear waste repository significantly. -- Abstract: Time evolution of criticality and burnup grades of the PBMR were investigated for reactor grade plutonium and minor actinides in the spent fuel of light water reactors (LWRs) mixed with thoria. The calculations were performed by employing the computer codes MCNP and MONTEBURNS 2.0 and using the ENDF/B-V nuclear data library. Firstly, the plutonium–thorium and minor actinides–thorium ratio was determined by using the initial k eff value of the original uranium fuel design. After the selection of the plutonium/minor actinides–thorium mixture ratio, the time-dependent neutronic behavior of the reactor grade plutonium and minor actinides and original fuels in a PBMR-400 reactor was calculated by using the MCNP code. Finally, k eff , burnup and operation time values of the fuels were compared. The core effective multiplication factor (k eff ) for the original fuel which has 9.6 wt.% enriched uranium was computed as 1.2395. Corresponding to this k eff value the reactor grade plutonium/thorium and minor actinide/thorium oxide mixtures were found to be 30%/70% and 50%/50%, respectively. The core lives for the original, the reactor grade plutonium/thorium and the minor actinide/thorium fuels were calculated as ∼3.2, ∼6.5 and ∼5.5 years, whereas, the corresponding burnups came out to be 99,000, ∼190,000 and ∼166,000 MWD/T, respectively, for an end of life k eff set equal to 1.02.

  19. Verification of Serpent code for the fuel analysis of a PBMR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Francois L, J. L.

    2015-09-01

    In this paper the models and simulations with the Monte Carlo code Serpent are presented, as well as the obtained results of the different analyzed cases in order to verify the suitability or reliability of the use of this code to ensure favorable results in the realization of a neutronic analysis of fuel for a Pebble Bed Modular Reactor (PBMR). Comparisons were made with the results reported in a report by the OECD/Nea relative to a high temperature reactor of spheres bed with plutonium reactor grade as fuel. The results show that the use of Serpent is appropriate, as these results are comparable with those reported in the report. (Author)

  20. Optimisation of Deep Burn Incineration of Reactor Waste Plutonium in a PBMR DPP-400 core

    International Nuclear Information System (INIS)

    Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik

    2013-01-01

    The incineration of pure Pu in the PBMR-400, using the simulated fuel sphere geometries, are not recommended, since it violated the safety limits. The addition of MA to Pu(PWR) is not recommended since this substantially increased the mass of heavy metals to be disposed in the spent fuel. Therefore a redesign of both the reactor and fuel sphere geometries are recommended, in order to resolve the safety issues

  1. RA Research reactor, Annual report 1986; Istrazivacki nuklearni reaktor RA - Izvestaj za 1986. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1986-12-01

    Research reactor RA Annual report for year 1985 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1986 godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA.

  2. Evaluation of design, leak monitoring, dnd NDEA strategies to assure PBMR Helium pressure boundary reliability - HTR2008-58037

    International Nuclear Information System (INIS)

    Fleming, K. N.; Smit, K.

    2008-01-01

    This paper discusses the reliability and integrity management (RIM) strategies that have been applied in the design of the PBMR passive metallic components for the helium pressure boundary (HPB) to meet reliability targets and to evaluate what combination of strategies are needed to meet the targets. The strategies considered include deterministic design strategies to reduce or eliminate the potential for specific damage mechanisms, use of an on-line leak monitoring system and associated design provisions that provide a high degree of leak detection reliability, and periodic nondestructive examinations combined with repair and replacement strategies to reduce the probability that degradation would lead to pipe ruptures. The PBMR RIM program for passive metallic piping components uses a leak-before-break philosophy. A Markov model developed for use in LWR risk-informed in-service inspection evaluations was applied to investigate the impact of alternative RIM strategies and plant age assumptions on the pipe rupture frequencies as a function of rupture size. Some key results of this investigation are presented in this paper. (authors)

  3. RA research nuclear reactor - Annual report 1985; Istrazivacki nuklearni reaktor RA - Izvestaj za 1985. godinu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-12-01

    Research reactor RA Annual report for year 1985 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1985 godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA.

  4. DORT-TD/THERMIX solutions for the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

    International Nuclear Information System (INIS)

    Tyobeka, Bismark; Pautz, Andreas; Ivanov, Kostadin

    2008-01-01

    In new reactor designs that are still under review such as the PBMR, not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400 MW OECD/NEA/NSC coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate the transient scenarios in the above-mentioned benchmark problem. Steady-state calculations results are compared with selected participants' results as well as transient models in which the diffusion and transport theory solutions of the same code system are directly compared. Several sensitivity studies are also shown in order to determine how much the change in certain parameters influences the overall behaviour of a given transient. It is shown in this paper that DORT-TD/THERMIX is a versatile tool which can be deployed for design and safety analyses of high temperature reactors of pebble-bed type. (authors)

  5. Separate effects tests to determine the effective thermal conductivity in the PBMR HTTU test facility

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, P.G., E-mail: pgr@mtechindustrial.com [School of Mechanical and Nuclear Engineering, North-West University, Private Bag X6001, Potchefstroom 2520 (South Africa); Toit, C.G. du; Antwerpen, W. van [School of Mechanical and Nuclear Engineering, North-West University, Private Bag X6001, Potchefstroom 2520 (South Africa); Antwerpen, H.J. van [M-Tech Industrial (Pty) Ltd., PO Box 19855, Noordbrug 2522 (South Africa)

    2014-05-01

    Thermal-fluid simulations are used extensively to predict the maximum fuel temperatures, flows, pressure drops and thermal capacitance of pebble bed gas cooled reactors in support of the reactor safety case. The PBMR company developed the HTTU non-nuclear test facility in cooperation with M-Tech Industrial (Pty) Ltd. and the North-West University in South Africa to conduct comprehensive separate effects tests as well as integrated effects tests to study the different thermal-fluid phenomena. This paper describes the separate effects tests that were conducted to determine the effective thermal conductivity through the pebble bed under near-vacuum conditions and temperatures up to 1200 °C. It also presents the measured temperature distributions and the methodology applied in the data analysis to derive the resultant values of effective thermal conductivity and its associated uncertainty.

  6. Separate effects tests to determine the effective thermal conductivity in the PBMR HTTU test facility

    International Nuclear Information System (INIS)

    Rousseau, P.G.; Toit, C.G. du; Antwerpen, W. van; Antwerpen, H.J. van

    2014-01-01

    Thermal-fluid simulations are used extensively to predict the maximum fuel temperatures, flows, pressure drops and thermal capacitance of pebble bed gas cooled reactors in support of the reactor safety case. The PBMR company developed the HTTU non-nuclear test facility in cooperation with M-Tech Industrial (Pty) Ltd. and the North-West University in South Africa to conduct comprehensive separate effects tests as well as integrated effects tests to study the different thermal-fluid phenomena. This paper describes the separate effects tests that were conducted to determine the effective thermal conductivity through the pebble bed under near-vacuum conditions and temperatures up to 1200 °C. It also presents the measured temperature distributions and the methodology applied in the data analysis to derive the resultant values of effective thermal conductivity and its associated uncertainty

  7. Development of a strategy for the management of PBMR spent fuel in South Africa

    Energy Technology Data Exchange (ETDEWEB)

    Smith, S.W., E-mail: Schalk.Smith@necsa.co.z [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa); Bredell, P.J. [South African Nuclear Energy Corporation Ltd (Necsa), Pretoria 0001 (South Africa)

    2010-10-15

    South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would form an integral part of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development.

  8. The coupled code system DORT-TD/THERMIX and its application to the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

    International Nuclear Information System (INIS)

    Pautz, A.; Tyobeka, B.; Ivanov, K.

    2009-01-01

    In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)

  9. The PBMR fuel plant: Proven technology in an advanced safety environment

    International Nuclear Information System (INIS)

    Braehler, G.; Froschauer, K.; Welbers, P.; Boyes, D.

    2008-01-01

    The PBMR Fuel Plant (PFP), to be constructed at the Pelindaba site near Johannesburg will fuel the first South African Pebble Bed Modular Reactor. The qualification of the PBMR fuel shall be based on past experience with fuel which was produced in the German NUKEM/HOBEG plant and irradiated in the German AVR reactor. Accordingly, the PFP must produce the same fuel as the German plant did, and consequently, the design of the PFP has in essence to be a copy of the NUKEM/HOBEG plant. As a reminder this plant had been operated in accordance with the German regulatory rules which were defined in the years 1970/80. Since then, the requirements with regard to radiological protection, criticality safety and emission control have been significantly tightened, and of course the PFP must be designed in accordance with the most advanced international norms and standards. The implications which follow from these two potentially conflicting requirements, as defined above, are highlighted, and technical solutions are presented. Hence, the change from administrative criticality safety control to technical control, i.e. the application of safe geometry as far as possible. and the introduction of technical solutions for the remaining safe mass regime will be described. A lot of equipment in the Kernel area and in the recycling areas needed to be redesigned in safe geometry. The sensitive processes for Kernel Calcining, for the Coating and the Over-coating remain under safe mass regime, but the safety against criticality is completely independent from staff activities and based on technical measures. A new concept for safe storage of large volumes of Uranium-containing liquids has been developed. Also, the change from relatively open handling of Uranium to the application of containment enclosures wherever release of radioactivity into the room atmosphere is possible, will be addressed. This change required redesign of all process steps requiring the handling of dry Uranium oxides

  10. Separate effects tests to determine the thermal dispersion in structured pebble beds in the PBMR HPTU test facility

    Energy Technology Data Exchange (ETDEWEB)

    Toit, C.G. du, E-mail: jat.dutoit@nwu.ac.za; Rousseau, P.G.; Kgame, T.L.

    2014-05-01

    Thermal-fluid simulations are used extensively to predict the maximum fuel temperatures, flows, pressure drops and thermal capacitance of pebble bed gas cooled reactors in support of the reactor safety case. The PBMR company developed the HTTF test facility in cooperation with M-Tech Industrial (Pty) Ltd. and the North-West University in South Africa to conduct comprehensive separate effects tests as well as integrated effects tests to study the different thermal-fluid phenomena. This paper describes the separate effects tests that were conducted to determine the effect of the porous structure on the fluid effective thermal conductivity due to the thermal dispersion. It also presents the methodology applied in the data analysis to derive the resultant values of the effective thermal conductivity and its associated uncertainty.

  11. Isotopic composition of precipitation at the station Ljubljana (Reaktor, Slovenia – period 2007–2010

    Directory of Open Access Journals (Sweden)

    Polona Vreča

    2014-12-01

    Full Text Available The stable isotopic composition of hydrogen and oxygen (δ2H and δ18O and the tritium activity (A were monitored in monthly collected precipitation at Ljubljana (Reaktor during the period 2007–2010. Monthly and yearly isotope variations are discussed and compared with those observed over the period 1981–2006 and with the basic meteorological parameters for Ljubljana (Bežigrad and Ljubljana (Hrastje stations for the period 2007−2010. The mean values for δ2H and δ18O, weighted by precipitation amount at Ljubljana (Reaktor, are –59.4 ‰ and –8.71 ‰. The reduced major axis local meteoric water line (LMWLRMA is δ2H = (8.19 ± 0.22×δ18O + (11.52 ± 1.97, while the precipitation weighted least square regression results in LMWLPWLSR-Re δ2H = (7.94 ± 0.21×δ18O + (9.76 ± 1.93. The lack of significant difference in the LMWL slopes indicates a relatively homogeneous distribution of monthly precipitation as well as the small number of low-amount monthly precipitation events with low deuterium excess. The deuterium excess weighted mean value is 10.3 ‰ which indicates the prevailing influence of the Atlantic air masses. The temperature coefficient of δ18O is 0.30 ‰/°C. Tritium activity in monthly precipitation shows typical seasonal variations, with a weighted mean tritium activity in this period of 8.5 TU. No decrease of mean annual activity is observed.

  12. RA Research reactor, Annual report 1970 - Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1970. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1970-12-15

    the pool. Refuelling was done three times by 'mixing' the irradiated fuel slugs with 5 to 6 fresh ones. Total exposure of the staff was increased due to activities during the shut-down period. Individual maximum exposure dose was not higher than 3.5 R. Cooperation with the French partner, concerned with the possibility of using 93% enriched fuel instead of the present low enriched, was continued. This would enable achievement of higher neutron flux, order of magnitude of 10{sup 14} n/cm{sup 2} sec. It is mentioned that there is doubt whether the plan for 1971 could be fulfilled for financial difficulties. [Serbo-Croat] Reaktor RA je u 1970. godini radio na nominalnoj snazi 160 dana i 40 dana na manjim snagama. Ukupni rad iznosio je 25968 MWh odnosno 3.87% vise od planiranog. Plan rada razlikovao se od prethodnih godina zbog slanja teske vode na rekoncentraciju u Francusku. Izotopski sastav teske vode je opao na 99.05% a posle rekoncentracije iznosi 99,96%. Odstupanje od plana rada u septembru mesecu bilo je izazvano kasnjenjem prispeca teske vode usled administrativnih teskoca i transporta. Pocetak kampanje odlozen je i zbog posledica ostecenja kosuljice jednog gorivnog elementa, sto se dogodilo odmah po pocetku rada pa je reaktor bio zaustavljen. U oktobru i novembru reaktor je radio 28 odnosno 25 dana respektivno sto je omogucilo da se nadoknadi izgubljeno vreme. Reaktor je koriscen za ozracivanja i eksperimente za 390 korisnika od cega 340 iz Instituta i 50 za korisnika izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. Zakljucuje se da je reaktor radio uspesno prema planu rada. Kracih zastoja u radu bilo je samo zbog teskoca sa cevovodima tehnicke vode kao posledica klizanja zemljista. Reaktor je samo dva puta sigurnosno zaustavljan zbog neispravnosti opreme odnosno laznog signala elektronske aparature sistema upravljanja. Vreme dok reaktor nije radio iskorisceno je za inspekciju unutrasnjosti reaktorskog suda. Pomocu

  13. RA Research reactor, Annual report 1971; Istrazivacki nuklearni reaktor RA - Izvestaj za 1971. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1971-12-15

    During 1971, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and 50 days at lower power levels. Total production mounted to 31606 MWh which is 5.3% higher than planned, and the highest annual level since the reactor started operation. Reactor was used for irradiation and experiments according to the demand of 425 users, of which 370 from the Institute and 55 external users. This report contains detailed data about reactor power and experiments performed in 1971. Discrepancies from the action plan, meaning higher production was achieved in June and December due to special demand of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to power cuts during reactor operation. There was no longer interruption caused by failures of the equipment. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. Shorter interruptions resulted from breaking of connectors in the technical water pipe system caused by soil sliding near the pumping station on the Danube. Total personnel exposure dose was lower than during previous years. There was no accident nor any event that could be called accidental. Decontamination od surfaces was less than during previous years. It was concluded that the successful operation in 1971 resulted from efficient work during past years. But, some of the activities were interrupted due to undefined policy concerned with operation of the RA reactor and financial issues. This involves study of the possibility to use highly enriched fuel that would increase the useful neutron flux and the reactor compatibility with similar reactors for the future ten years. Another project that has been interrupted is construction of the emergency core cooling system which is important for the reactor safety. Financial problems have influenced not only the reactor operation but the number of employees, which could cause negative

  14. Disain Sistem Pemantauan Lingkungan Untuk Evaluasi Lepasan Radionuklida dari Subsistem pada Kecelakaan Reaktor Daya PWR

    Directory of Open Access Journals (Sweden)

    Sri Kuntjoro

    2013-03-01

    Full Text Available PLTN. (Pembangkit Listrik Tenaga Nuklir sebagai sumber energi baru dipilih sebagai alternatif, karena memiliki berbagai kelebihan yaitu ramah lingkungan, pasokan bahan bakar yang tidak bergantung musim, serta harganya yang dapat bersaing dengan pembangkit listrik yang lain. Namun demikian, adanya keraguan sebagian masyarakat tentang keselamatan radiasi PLTN, maka pemerintah harus bisa meyakinkan tentang operasi PLTN yang aman dan selamat. Penelitian tentang disain sistem pemantauan lingkungan untuk evaluasi lepasan radionuklida dari subsistem reaktor dan lingkungan akibat terjadinya kecelakaan pada reaktor daya telah dilakukan. Penelitian dilakukan dengan melakukan perhitungan sebaran radionuklida ke subsistem dan lingkungan serta membuat sistim monitoring radiasi di lingkungan. Sistem monitoring lingkungan terdiri dari system pencacah radiasi, sistem peringatan dini, sistem pengukuran meteorologi, sistem GPS dan system GIS. Sistem pencacah radiasi digunakan untuk mencatat data radiasi, sistem pengukuran meteorologi digunakan untuk mencatat data arah dan kecepatan angin, sedangkan sistem GPS digunakan untuk menentukan data posisi pengukuran. Data tersebut kemudian dikirimkan ke system akuisisi data untuk ditransmisikan ke pusat kendali. Pengumpulan dan pengiriman data dilakukan melalui SMS menggunakan perangkat modem yang ditempatkan di ruang kendali. Ruang kendali menerima data dari berbagai tempat pengukuran. Dalam hal ini ruang kendali memiliki fungsi sebagai SMS gateway. Sistem ini dapat memvisualisasi untuk lokasi pengukuran yang berbeda. Selanjutnya, data posisi dan data radiasi diintegrasikan dengan peta digital. Integrasi sistem tersebut kemudian divisualisasikan dalam personal komputer. Untuk posisi pengukuran terlihat langsung di peta dan untuk data radiasi ditampilkan di monitor dengan tanda lingkaran merah atau hijau yang digunakan sebagai pemonitor batas aman radiasi. Bila tanda lingkaran berwarna merah maka akan menyalakan alarm di

  15. Development of a strategy for the management of PBMR spent fuel in South Africa - HTR2008-58047

    International Nuclear Information System (INIS)

    Smith, S. W.; Bredell, P. J.; Meyer, W. C. M. H.

    2008-01-01

    South Africa is planning to expand its nuclear power generating capacity by deploying a number of pressurized-water reactors and pebble-bed modular reactors. It can be expected that this program will impact on the current and planned spent fuel and radioactive waste management systems in South Africa. This paper proposes an approach to develop a strategy for the management of PBMR spent fuel that would contribute to the optimization of the overall national radwaste management system. The approach is expected to provide a conceptual spent fuel management strategy and will also highlight areas that need to be further developed, thus providing guidance for basic technology development. (authors)

  16. PEMBUATAN SERBUK U-6Zr DENGAN PENGKAYAAN URANIUM 19,75 % UNTUK BAHAN BAKAR REAKTOR RISET

    Directory of Open Access Journals (Sweden)

    Masrukan Masrukan

    2016-03-01

    Full Text Available ABSTRAK PEMBUATAN SERBUK PADUAN U-6Zr DENGAN PENGKAYAAN URANIUM 19,75 % UNTUK BAHAN BAKAR REAKTOR RISET. Telah dilakukan pembuatan serbuk paduan U-6Zr dengan pengkayaan 19,75 % untuk bahan bakar reaktor riset. Pembuatan bahan bakar U-6Zr ini dalam rangka mencari bahan bakar baru yang mempunyai densitas tinggi untuk mengganti bahan bakar yang sudah ada U3Si2-Al. Tujuan dari percobaan ini untuk mengetahui sifat-sifat serbuk paduan U- 6Zr yang diperoleh dari proses hydriding-dehydriding sebagai kandidat bahan bakar reaktor riset. Serbuk yang diperoleh dari proses hydriding-dehydriding dikenai pengujian, diantaranya pungujian komposisi kimia, densitas, kandungan hidrogen, fasa dan sifat termal. Hasil pengujian komposisi kimia menunjukkan beberapa unsur seperti Al, Ca, Cu, dan Ni melebihi batas yang diijinkan dimana masing-masing unsur terdapat sebesar 202,21 ppm; 214,05 ppm; 61,25 ppm dan 134,53 ppm. Pada pengujian diperolah densitas serbuk U-6Zr sebesar 13,58 g/cm3 dan pada pengujian kandungan hidrogen sisa diperoleh kandungan hidrogen sebesar 0,16 %. Untuk pengujian fasa, diperoleh fasa αU dan δU, sedangkan pada pengujian sifat termal yakni transformasi temperatur terdapat dua puncak yakni puncak pertama terjadi pada temperatur 274 hingga 311 oC dan puncak kedua terjadi pada temperatur 493 hingga 527oC. Puncak pertama terjadi reaksi endotermik dengan menyerap panas sebesar ∆H = 6,23 cal/g tetapi tidak terbentuk fasa baru, sedangkan puncak kedua terjadi reaksi eksotermik dengan mengeluarkan panas sebesar ∆H = -9.34 cal/g dan terbentuk fasa αZr. Sementara itu, dari pengujian kapasitas panas pada temperatur 34 hingga 75 oC, terjadinya penurunan nilai kapasitas panas yang disertai dengan penyerapan panas. Pada temperatur yang lebih tinggi hingga temperatur 437oC nilai kapasitas panas menjadi lebih kecil disertai pengeluaran panas. Reaksi termokimia antara Zr dengan hidrogen sisa menunjukkan terbentuknya fasa αZr yang diindikasikan oleh reaksi

  17. Studi Model Benchmark Mcnp6 Dalam Perhitungan Reaktivitas Batang Kendali Htr-10

    OpenAIRE

    Jupiter S.Pane, Zuhair, Suwoto, Putranto Ilham Yazid

    2016-01-01

    STUDI MODEL BENCHMARK MCNP6 DALAM PERHITUNGAN REAKTIVITAS BATANG KENDALI HTR-10. Dalam operasi reaktor nuklir, sistem batang kendali memainkan peranan yang sangat penting karena didesain untuk mengendalikan reaktivitas teras dan memadamkan reaktor. Nilai reaktivitas batang kendali harus diprediksi secara akurat melalui eksperimen dan perhitungan. Makalah ini mendiskusikan model Benchmark dalam perhitungan reaktivitas batang kendali reaktor HTR-10. Perhitungan dikerjakan dengan program transpo...

  18. Helium storage and control system for the PBMR

    International Nuclear Information System (INIS)

    Verkerk, E.C.

    1997-01-01

    The power conversion unit will convert the heat energy in the reactor core to electrical power. The direct-closed cycle recuperated Brayton Cycle employed for this concept consists of a primary helium cycle with helium powered turbo compressors and a power turbine. The helium is actively cooled with water before the compression stages. A recuperator is used to preheat the helium before entering the core. The start of the direct cycle is initiated by a mass flow from the helium inventory and control system via a jet pump. When the PBMR is connected to the grid, changes in power demand can be followed by changing the helium flow and pressure inside the primary loop. Small rapid adjustments can be performed without changing the helium inventory of the primary loop. The stator blade settings on the turbines and compressors are adjustable and it is possible to bypass reactor and turbine. This temporarily reduces the efficiency at which the power conversion unit is operating. Larger or long term adjustments require storage or addition of helium in order to maintain a sufficient level of efficiency in the power conversion unit. The helium will be temporarily stored in high pressure tanks. After a rise in power demand it will be injected back into the system. Some possibilities how to store the helium are presented in this paper. The change of helium inventory will cause transients in the primary helium loop in order to acquire the desired power level. At this stage, it seems that the change of helium inventory does not strongly effect the stability of the power conversion unit. (author)

  19. RA Research reactor, Annual report 1969; Istrazivacki nuklearni reaktor RA - Izvestaj za 1969. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1969-12-15

    years of operation as well as purification of heavy water. [Serbo-Croat] Reaktor RA je u 1969. godini radio na nominalnoj snazi 200 dana i 15 dana na manjim snagama. Ukupni rad iznosio je 31131 MWh odnosno 3.77% vise od planiranog. Reaktor je koriscen za ozracivanja i eksperimenta prema zahtevima 463 korisnika iz Instituta i 63 izvan Instituta. Ovaj izvestak sadrzi detaljne podatke o snazi na kojoj je reaktor radio tokom 1969. godine i o uradjenim eksperimentima. Zakljuceno je da je reaktor RA uspesno radio prema planu rada. Da nije bilo problema sa napajanjem elektricnom energijom tokom poslednja tri meseca i niskog vodostaja Dunava u septembru i oktobru protekle godine bila bi to najuspesnija godina od pustanja reaktora u pogon. Broj sigurnosnih zaustavljanja nije bio veci u odnosu na prethodne dve godine i pored poteskoca u poslednjem kvartalu. Osoblje je bilo izlozeno povecanim dozama zracenja usled tri incidenta. Jedan je bio raspadanje kenera sa srebrom (zbog duzeg stajanja u aktivnoj zoni), sto je uzrokovalo kontaminaciju radne platforme, tako da je fon porastao za 10 do 100 puta od normalnohg Druga dva slucaja su bila otkazivanje uredjaja za mesanje goriva u tehnoloskim kanalima. Zamena goriva je radjena cetiri puta u toku godine, utroseno je 499 svezih gorivnih elemenata. Primenjena je metoda mesanja svezih gorivnih elemenata sa koriscenim gorivnim elementima u gorivnom kanalu. Dekontaminacija povrsina bila je na nivou prethodnih godina i pored problema sa srebrom. Kako su sa reaktora tokom godine otisla dva saradnika sa visokom spremom broj ljudi je opao na neophodan minimum za pogon i odrzavanje reaktora. Navrsavajuci u ovoj godini deset godina rada moze se reci da su rad i stanje opreme na tehnicki solidnom nivou. Kako bi se posle deset godina rada izvrsila kontrola vaznih komponenti reaktora i obavila rekoncentracija teske vode, za 1970. godinu je planirano da se proizvodnja smanji na 25000 MWh, a baziran je na istim principima kao i planovi za prethodne

  20. Pengaruh Laju Alir Inlet Reaktor MSL terhadap Reduksi BOD, COD, TSS, dan Minyak/Lemak Limbah Cair Industri Minyak Goreng

    Directory of Open Access Journals (Sweden)

    Salmariza Sy

    2017-06-01

    Full Text Available This research was conducted by treating edible oil industry wastewater used Multi Soil Layering (MSL method. The MSL reactor was built from a 200x120x200 cm concrete basin. Andisol soil was mixed with sawdust and fine charcoal at each ratio 5:1:1 based on dry weight as an impermeable layer. The flow rate variations were 250, 500, 1000, and 1500 L/m2.day. The observed pollutant parameters were BOD, COD, TSS, oil/fat, and pH. The results showed that MSL reactor was effective to decrease the pollutant content of edible oil industry wastewater. The reactor could reduce concentration of effluent parameters below standard except for oil/fat parameters at high flow rates. In the effluent was found BOD 0.66-14.22 mg/L, COD 5-69 mg/L, TSS 9-26 mg/L, and oil/fat 2-9 mg/L. The flow rate had an effect on reduction efficiency of BOD, COD, TSS, and oil/fat but did not effect pH as all flow rate could raise pH 6.37-6.95 became pH 6.99-7.24. The lower the flow rate the higher the reduction efficiency. The reduction efficiency at flow rates 250 and 1500 L/m2 days for BOD were 99% and 86%, COD were 96% and 71%, TSS were 88% and 77%, and oil/fat were 80% and 60%.ABSTRAK  Penelitian ini dilakukan dengan mengolah air limbah industri minyak goreng menggunakan metoda Multi Soil Layering (MSL. Reaktor MSL dibuat dari beton berbentuk bak ukuran 200x120x200 cm. Tanah andisol dicampur dengan serbuk gergaji dan arang halus pada rasio masing-masing 5:1:1 berdasarkan berat kering sebagai penyusun lapisan impermeable. Variasi laju alir yaitu 250, 500, 1000, dan 1500 L/m2.hari. Parameter pencemar yang dianalisis meliputi BOD, COD, TSS, minyak/lemak, dan pH. Hasil penelitian menunjukkan bahwa reaktor MSL sangat efektif untuk menurunkan kandungan zat pencemar limbah cair industri minyak goreng. Reaktor dapat mereduksi konsentrasi parameter outlet sampai dibawah baku mutu yang distandarkan kecuali untuk parameter miyak/lemak pada perlakuan laju alir tinggi. Pada effluen

  1. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  2. PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-04-01

    Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

  3. RA Research reactor, Annual report 1968 - Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1968. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-12-15

    During 1968, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and during 50 days at lower power levels. Total production amounted to 31051 MWh which is 3.5% higher than planned. reactor was used for irradiation and experiments according to the demand of 600 users, of which 517 from the Institute and 83 externals users. This report contains detailed data about reactor power and experiments performed in 1968. It is concluded that the reactor operation was more successful than during previous years. There was only one longer interruption which lasted 27 hours because of the power cut on the cable for the pump station on Danube. Number of safety shutdowns were at the same level as during last year. The only significant incident in 1968 was air contamination with the radioactive argon in the reactor hall. The reactor operation was not interrupted although the hall was evacuated for two hours. The was no significant exposure of the staff. In April and September the integral dosed were higher than during other months because of the accident during refueling (mixing the slugs with irradiated and fresh fuel). There was no significant surface contamination, i.e. the decontaminated surface were negligible. Due to 'mixing' refueling scheme. [Serbo-Croat] Reaktor RA je u 1968. godini radio na nominalnoj snazi od 6,5 MW 190 dana i 50 dana na manjim snagama. Ukupni rad iznosio je 31051 MWh odnosno 3,5% vise od planiranog. Reaktor je koriscen za ozracivanja i eksperimente za 600 korisnika od cega 517 iz Instituta i 83 za korisnika izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. Zakljucuje se da je reaktor radio uspesnije nego prethodnih godina. U toku 1968. godine samo je jedan duzi prekid u radu od 27 casova izazvan zbog proboja kablovske glave na odvodu za pumpnu stanicu na Dunavu. Sigurnosna zaustavljanja bila su na proslogodisnjem nivou. Jedini znacajniji incident u 1968. godini, bio je kontaminacija vazduha

  4. Conceptual design of a nucleo electric simulator with PBMR reactor based in Reduced order models; Diseno conceptual de un simulador de nucleo electrica con reactor PBMR basado en modelos de orden reducido

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: jms0620@yahoo.com

    2005-07-01

    This project has as purpose to know to depth the operation of a PBMR nucleo electric type (Pebble Bed Modular Reactor), which has a reactor of moderate graphite spheres and fuel of uranium dioxide cooled with Helium and Brayton thermodynamic cycle. The simulator seeks to describe the dynamics of the one process of energy generation in the nuclear fuel, the process of transport toward the coolant one and the conversion to mechanical energy in the turbo-generators as well as in the heat exchangers indispensable for the process. The dynamics of reload of the fuel elements it is not modeled in detail but their effects are represented in the parameters of the pattern. They are modeled also the turbo-compressors of the primary circuit of the work fluid. The control of the power of the nuclear reactor is modeled by means of reactivity functions specified in the simulation platform. The proposed mathematical models will be settled in the platform of simulation of Simulink-Mat Lab. The proposed control panels for this simulator can be designed and to implement using the box of tools of Simulink that facilitates this process. The work presents the mathematical models more important used for their future implementation in Simulink. (Author)

  5. PENGARUH KONSENTRASI ZrO2 TERHADAP KORELASI PERPINDAHAN PANAS NANOFLUIDA AIR-ZrO2 UNTUK PENDINGIN REAKTOR

    Directory of Open Access Journals (Sweden)

    K.A. Sudjatmi

    2015-03-01

    Full Text Available Sejalan dengan perkembangan konsep keselamatan pasif pada sistem keselamatan PLTN, maka sistem perpindahan panas konveksi alam memegang peranan penting. Pemakaian nanofluid sebagai fluida pendingin pada sistem keselamatan nuklir dapat digunakan pada Sistem Pendingin Teras Darurat dan Sistem Pendingin Pengungkung Luar Reaktor. Beberapa peneliti telah melakukan studi desain konseptual aplikasi nanofluid untuk meningkatkan keselamatan AP1000 dan sistem pendingin teras darurat pada reaktor daya eksperimen. Penerapan nanofluida juga mulai dikembangkan melalui hasil penelitian perpindahan panas konveksi alamiah pada sub-buluh dengan nanofluida sebagai fluida kerjanya sangat dibutuhkan. Penelitian ini bertujuan untuk menentukan pengaruh perubahan konsentrasi ZrO2 terhadap korelasi perpindahan panas konveksi alamiah dengan pendekatan eksperimental. Data eksperimental yang diperoleh digunakan untuk mengembangkan korelasi umum empirik perpindahan panas konveksi alamiah. Metode penelitian dengan menggunakan alat uji sub-buluh vertikal dengan geometri segitiga dan segiempat menggunakan air dan nanofluida air-ZrO2 sebagai fluida kerjanya. Konsentrasi nanopartikel dalam larutan yang digunakan sebesar 0,05 %, 0,10% dan 0,15 % dalam persen berat. Hasil penelitian menunjukan bahwa untuk bilangan Rayleigh yang sama, kemampuan pemindahan kalor oleh nanofluida air-ZrO2 lebih baik dari pada pemindahan kalor oleh air. Namun peningkatan konsentrasi nanofluida tidak selalu mendapatkan kemampuan pemindahan kalor yang lebih baik. Kata kunci: nanofluida air-ZrO2, konveksi alamiah, sub-buluh segitiga, sub-buluh segi segiempat   In line with the development of the passive safety concept for the safety systems of nuclear power plants, the natural convection heat transfer system plays an important role. The nanofluid as coolant fluid on nuclear safety system can be used in Emergency core cooling system and in reactor coolant system confinement. Several researchers have

  6. RA Research reactor, Annual report 1988; Istrazivacki nuklearni reaktor RA - Izvestaj za 1988. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1988-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1989. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3705/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjering RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  7. PERFORMA OKSIDASI METAN PADA REAKTOR KONTINYU DENGAN PENINGKATAN KETEBALAN LAPISAN BIOCOVER LANDFILL

    Directory of Open Access Journals (Sweden)

    Opy Kurniasari

    2013-11-01

    Full Text Available PERFORMANCE OF METHANE OXIDATION IN CONTINUOUS REACTOR BY BIOCOVER LANDFILL FILM THICKNESS IMPROVEMENT. Municipal solid waste (MSW handling in Indonesia is currently highly dependent on landfilling at the final disposal facility (TPA, which generally operated in layer-by-layer basis, allowing the anaerobic (absent of oxygen process. This condition will certainly generate biogas in the form of methane (CH4 and CO2. Methane is a greenhouse gas with a global warming potential greater than CO2, and can absorb infrared radiation 23 times more efficient than CO2 in the period of over 100 years. One way that can be done to reduce methane gas from landfills that escape into nature is to oxidize methane by utilizing landfill cover material (biocover as methane-oxidizing microorganism media. Application of compost as landfill cover material is a low-cost approach to reduce emissions so are suitable for developing countries. The compost used in this study was compost landfill mining, which is degraded naturally in landfill. The purpose of this study was to evaluate the ability of biocover to oxidize the methane on a certain layer thickness with a continuous flow conditions. Three column reactors were used, which were made of flexy glass measuring 70 cm in high and 15 cm in diameter. The methane flowed from the bottom of the reactor continuously at a flow rate of 5 ml/minute. The columns were filled with biocover compost landfill mining with layer thickness of 5, 25, 35 and 60 cm. The results showed that the thicker layer of biocover, the higher the efficiency of methane oxidation. The oxidation efficiency obtained in each layer thickness of 15, 25, 35 and 60 cm was 56.43%, 63.69%, 74.58% and 80, 03% respectively, with the rate of oxidation of 0.29 mol m-2 d-1 and the fraction of oxidation of 99%. The oxidation result was supported by the identification of bacteria isolated in this experiment, namely metanotrophic bacteria that have the ability to oxidize

  8. Conceptual design of a nucleo electric simulator with PBMR reactor based in Reduced order models

    International Nuclear Information System (INIS)

    Valle H, J.; Morales S, J.B.

    2005-01-01

    This project has as purpose to know to depth the operation of a PBMR nucleo electric type (Pebble Bed Modular Reactor), which has a reactor of moderate graphite spheres and fuel of uranium dioxide cooled with Helium and Brayton thermodynamic cycle. The simulator seeks to describe the dynamics of the one process of energy generation in the nuclear fuel, the process of transport toward the coolant one and the conversion to mechanical energy in the turbo-generators as well as in the heat exchangers indispensable for the process. The dynamics of reload of the fuel elements it is not modeled in detail but their effects are represented in the parameters of the pattern. They are modeled also the turbo-compressors of the primary circuit of the work fluid. The control of the power of the nuclear reactor is modeled by means of reactivity functions specified in the simulation platform. The proposed mathematical models will be settled in the platform of simulation of Simulink-Mat Lab. The proposed control panels for this simulator can be designed and to implement using the box of tools of Simulink that facilitates this process. The work presents the mathematical models more important used for their future implementation in Simulink. (Author)

  9. Research nuclear reactor RA - Annual Report 1989; Istrazivacki nuklearni reaktor RA - Izvestaj za 1989. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1989-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1989. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3705/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjering RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  10. Research nuclear reactor RA - Annual Report 1996; Istrazivacki nuklearni reaktor RA - Izvestaj za 1996. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1996-12-01

    ugovoru trebalo je da celokupna nova instrumentacija za reaktor RA bude isporucena Institutu u Vinci do kraja 1990. godine, ali je do septembra 1991. godine isporuceno svega 80% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju izrecena od strane ove organizacije Ujedinjenih nacija. Demontirana je postojeca instrumentacija u maksimalno mogucem obimu, kako bi se zadrzala neka osnovna merenja neophodna i u uslovima kada reaktor nije u pogonu. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Kontrola goriva od strane inspektora MAAE obavljana je jedanput mesecno. U ovom izvestajnom periodu na reaktoru RA bilo je zaposleno prosecno 43 radnika, sto je dovoljan broj u uslovima remontnih i investicionih radova. Nedostatak finansijskih sredstava za odrzavanje reaktora RA je neresen problem i u ovom periodu. Ovaj izvestaj podeljen je u dve celine: pogon i odrzavanje reaktora i zastita od zracenja na reaktoru RA.

  11. VERIFIKASI PAKET PROGRAM MVP-II DAN SRAC2006 PADA KASUS TERAS REAKTOR VERA BENCHMARK.

    Directory of Open Access Journals (Sweden)

    Jati Susilo

    2015-03-01

    Full Text Available Dalam penelitian ini dilakukan verifikasi perhitungan benchmark VERA pada kasus Zero Power Physical Test (ZPPT teras reaktor Watts Bar 1. Reaktor tersebut merupakan jenis PWR kelas 1000 MWe yang didesain oleh Westinghouse, tersusun dari 193 perangkat bahan bakar 17×17 dengan 3 jenis pengkayaan UO2 yaitu 2,1wt%, 2,619wt% dan 3,1wt%. Perhitungan nilai k-eff dan distribusi faktor daya dilakukan pada siklus operasi pertama teras dengan kondisi beginning of cycle (BOC dan hot zero power (HZP. Posisi batang kendali dibedakan menjadi uncontrolled (semua batang kendali berada di luar teras, dan controlled (batang kendali Bank D didalam teras. Paket program komputer yang digunakan dalam perhitungan adalah MVP-II dan SRAC2006 modul CITATION dengan data pustaka tampang lintang ENDF/B-VII.0. Hasil perhitungan menunjukkan bahwa perbedaan nilai k-eff teras pada kondisi controlled dan uncontrolled antara referensi dengan MVP-II (-0,07% dan -0,014% dan SRAC2006 (0,92% dan 0,99% sangat kecil atau masih dibawah 1%. Perbedaan faktor daya maksimum teras pada kondisi controlled dan uncontrolled dengan referensi dengan MVP-II adalah 0,38% dan 1,53%, sedangkan dengan SRAC2006 adalah 1,13% dan -2,45%. Dapat dikatakan bahwa kedua paket program komputer menunjukkan hasil perhitungan yang sesuai dengan nilai referensi. Dalam hal penentuan kekritisan teras, maka hasil perhitungan MVP-II lebih konservatif dibandingkan dengan SRAC2006. Kata kunci : MVP-II, SRAC2006, PWR, VERA   In this research, verification calculation for VERA core physics benchmark on the Zero Power Physical Test (ZPPT of the nuclear reactor Watts Bar 1. The reactor is a 1000 MWe class of PWR designed by Westinghouse, arranged from 193 unit of 17×17 fuel assembly consisting 3 type enrichment of UO2 that are 2.1wt%, 2.619wt% and 3.1wt%. Core power factor distribution and k-eff calculation has been done for the first cycle operation of the core at beginning of cycle (BOC and hot zero power (HZP. In this

  12. Studi Awal Desain Pebble Bed Reactor Berbasis Htr-pm Dengan Skema Resirkulasi Bahan Bakar Once-through-then-out

    OpenAIRE

    Setiadipura, Topan; Pane, Jupiter Sitorus; Zuhair, Zuhair

    2016-01-01

    STUDI AWAL DESAIN PEBBLE BED REACTOR BERBASIS HTR-PM DENGAN RESIRKULASI BAHAN BAKAR ONCE-THROUGH-THEN-OUT. Reaktor nuklir tipe pebble bed reactor (PBR) adalah salah satu reaktor canggih dengan fitur keselamatan pasif yang kuat. Pada desain tipe ini berpotensi untuk dilakukan kogenerasi yang bermanfaat untuk pengolahan berbagai mineral di berbagai pulau di Indonesia. Operasi PBR dapat lebih disederhanakan dengan menerapkan skema pengisian bahan bakar once-through-then-out (OTTO) dimana bahan b...

  13. RA Research nuclear reactor - Annual report 1987; Istrazivacki nuklearni reaktor RA - Izvestaj o radu za 1987. godinu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1987. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3509/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjjring RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  14. ANALISIS PENGARUH IRADIASI FLUENS NEUTRON CEPAT TERHADAP BERILIUM REFLEKTOR REAKTOR RSG-GAS

    Directory of Open Access Journals (Sweden)

    Sri Kuntjoro

    2015-04-01

    Full Text Available Telah dilakukan analisis iradiasi fluens neutron cepat terhadap berilium reflektor reaktor RSG-GAS. Analisis dilakukan dengan cara melakukan pengukuran fluks neutron di posisi berilium elemen dan berilium blok yang berfungsi sebagai reflector. Selanjutnya dilakukan perhitungan untuk menentukan apakah ada pengaruh fluens neutron selama berilium berada di teras reaktor. Selain cara tersebut dilakukan pula visualisasi untuk memastikan ada tidaknya deformasi pada berilium akibat iradiasi. Hasil pengukuran fluks dan fluens neutron cepat maksimal pada daya 200 kW untuk berilium elemen posisi E-2 sebesar 2,30E+07 n/cm2s dan 4,19E+17 n/cm2, J-8 sebesar 3,70E+07 n/cm2s dan 6,74E+17 n/cm2. Hasil pengukutan pada posisi B-3 sebesar 2,19E+12 n/cm2s dan 3,99E+22 n/cm2, G-10 sebesar 2,12E+12 n/cm2s dan 3,86E+22 n/cm2, serta berilium blok posisi (5-6 sebesar 5,02E+07 n/cm2s dan 9,15E+17 n/cm2, (C-D sebesar 2,32E+07 n/cm2s dan 4,23E+17 n/cm2. Deformasi yang diperoleh untuk berilium elemen (∆L/L posisi E-2 sebesar 1,12E-08, J-8 sebesar 1,84E-08, B-3 sebesar 1,60E-03, posisi G-10 sebesar 1,55E-03, sedangkan pada berilium blok di posisi 5-6 sebesar 2,52E-08 dan C-D sebesar 1,13E-08. Dari hasil ini disimpulkan tidak terjadi deformasi pada berilium elemen dan berilium blok. Hasil ini dibuktikan pula dari pengamatan visual, dimana tidak terlihat adanya deformasi pada berilium tersebut. Kata kunci : fluks, fluens, berilium elemen, berilium blok   Analysis of influence fast neutron fluence irradiated to the RSG-GAS beryllium reflector have been done. Methods of analysis was carried out by measuring fluxs neutron in beryllium element and block positio that function as reflector. The calculation done for determination it is there any influence of neutron as long as beryllium in the core. Bisede that, visualization done to make sure it there is any deformation at beryllium as efect of irradiation. Fluxs and fluences of beryllium element measurement result in 200 k

  15. RA Research reactor, Annual report 1973; Istrazivacki nuklearni reaktor RA - Izvestaj za 1973. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1973-12-15

    During 1973, RA reactor was operated at nominal power for 4687 hours and 54 hours at lower power levels. The total production was 30504 MWh which is 1.6% higher than planned. Practically there was no discrepancies from the plan, since the action plan was corrected at the beginning of this year caused by the demand of changing the schedule for refuelling for the purpose of 'power excursion' experiment. The reactor was used for irradiation and experiments according to the demand of 336 users. This report contains detailed data about reactor power and experiments performed in 1973. Total number safety shutdowns was 12, of which 7 were caused by power cuts. Three shutdowns caused by failures of the equipment were caused by failures of new electronic tubes. Two shutdowns were caused by the operators. There have been three shorter interruptions announced power cuts. Total personnel exposure dose was lower than during previous years. There were no accidents during this year. Decontamination of surfaces was less than during previous years. Practically there was no surface contamination, and the quantity of collected radioactive waste was lower than previously. There were no liquid radioactive effluents. It was concluded that the successful operation in 1973 has a special significance taking into account the financial crisis. There still remains a number of unsolved problems related to: completing the inventory of spare parts, exchange of some elements of the equipment, exchange of instrumentation, and purchase of the highly enriched fuel. [Serbo-Croat] Reaktor RA je u 1973. godini radio na nominalnoj snazi 4687 sati i 54 sata na manjim snagama. Ukupni rad iznosio je 30504 MWh odnosno 1,6% vise od planiranog. Prakticno nije bilo odstupanja od plana rada koji je pocetkom godine korigovan zbog promene planiranih izmena goriva usled izvodjenja eksperimenta 'ekskurzije snage'. Reaktor je koriscen za ozracivanja i eksperimente za 336 korisnika. Ovaj izvestaj sadrzi detaljne

  16. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA, Deo 1 - Pogon, odrzavanje i eksploatacija reaktora u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Milosevic, M; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1981-12-15

    biggest difficulty was maintenance of reactor instrumentation. During 1981 the reactor was operated safely, there was no accident nor incident that would affect the safety of reactor personnel or the environment. The testing operation will be continued in 1982,and the experience so far shows that the program would be successfully fulfilled on the whole. [Serbo-Croat] Nuklearni reaktor RA prestao je sa radom nakon martovske kampanje 1979. godine usled pojave talozenja oksihidrata aluminijuma na kosuljicama gorivnih elemenata. Odgovarajucim resenjima Sanitarnog inspektorata Republickog sekretarijata za zdravje i socijalnu politiku SR Srbije i generalnog direktora Instituta za nuklearne nauke 'Boris Kidric', Vinca zabranjen je dalji rad reaktora sve dok se ne utvrde uzroci stvaranja oksihidrata aluminijuma i njihovog talozenja, preduzmu mere za njihovo uklanjanje i ne obezbede potrebni uslovi za normalan nastavak rada reaktora. Do kraja 1979. i tokom 1980. godine, nakon niza izvrsenih analiza i utvrdjivanja uzroka koji su doveli do zaustavljanja rada reaktora, izvrsene su sve neophodne pripreme za ponovno pustanje reaktora u rad. Polazeci od cinjenice da na reaktoru RA ne postoji sistem za hladjenje jezgra u slucaju udesa i da ne postoji adekvatan sistem za filtriranje potencijalno zagadjenog vazduha, a saglasno sa novim propisima o pustanju u rad i probnom radu nuklearnih objekata, Sanitarni inspektorat je doneo privremeno resenje kojim se dozvoljava pustanje reaktora u rad, tj. izvodjenje tzv. 'nultog eksperimenta' uz ogranicenje snage na 1% od vrednosti nominalne snage. Na osnovu dobijene dozvole, reaktor RA je ponovo pusten u rad 21. januara 1981. godine, kada je dostignuta kriticnost sa jezgrom sastavljenim iskljucivo od gorivnih elemenata od 80% obogacenog uranijuma. Eksperiment je zavrsen krajem marta, nakon cega je zatrazena dozvola za probni rad na vecim snagama i potom za rad na punoj snazi. Uzimajuci postojece stanje reaktora RA doneto je resenje kojim se

  17. Research nuclear reactor RA - Annual Report 1991; Istrazivacki nuklearni reaktor RA - Izvestaj za 1991. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-01-01

    zahvat, koji se odnosi na zamenu celokupne instrumentacije je u toku, ali njegova realizacija u 1991. godini kasni zbog zastoja u isporuci opreme koja se izradjuje u Sovjetskom savezu. Izradu ove opreme finansira Medjunarodna agencija za atomsku energiju kroz ugovor sklopljen decembra 1988. godine sa moskovskom firmom Atomenergoexport. Prema tom ugovoru trebalo je da celokupna nova instrumentacija za reaktor RA bude isporucena Institutu u Vinci do kraja 1990. godine, ali je do septembra 1991. godine isporuceno svega 56% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju izrecena od strane ove organizacije Ujedinjenih nacija. U 1991. godini na reaktoru RA je demontirana postojeca instrumentacija u maksimalno mogucem obimu, kako bi se zadrzala neka osnovna merenja neophodna i u uslovima kada reaktor nije u pogonu. Pri kraju je izrada odredjenih konstrukcionih elemenata za novu instrumentaciju, koju je prihvatio da realizuje Institut u Vinci, kako bi se ubrzala realizacija ovog projekta. Ako sva predvijena oprema ne bude isporucena do kraja marta 1992. godine, nece moci da se otpocne sa probnim radom reaktora u prvoj polovini 1993. godine, kako je planirano. Na realizaciji projekta u 1991. godini ucestvovalo je efektivno 53 radnika, sto je dovoljan broj u uslovima remontnih i investicionih radova. Godisnji izvestaj o radu nuklearnog reaktora RA za 1992. godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA.

  18. Research nuclear reactor RA - Annual Report 1994; Istrazivacki nuklearni reaktor RA - Izvestaj za 1994. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1994-12-01

    pouzdan rad ovog reaktora. Poslednji, i ujedno najveci zahvat, koji se odnosi na zamenu celokupne instrumentacije je u toku, ali njegova realizacija u 1994. godini kasni zbog zastoja u isporuci opreme koja se izradjuje u Sovjetskom savezu. Izradu ove opreme finansira Medjunarodna agencija za atomsku energiju kroz ugovor sklopljen decembra 1988. godine sa moskovskom firmom Atomenergoexport. Prema tom ugovoru trebalo je da celokupna nova instrumentacija za reaktor RA bude isporucena Institutu u Vinci do kraja 1990. godine, ali je do septembra 1991. godine isporuceno svega 56% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju izrecena od strane ove organizacije Ujedinjenih nacija. U 1991. godini na reaktoru RA je demontirana postojeca instrumentacija u maksimalno mogucem obimu, kako bi se zadrzala neka osnovna merenja neophodna i u uslovima kada reaktor nije u pogonu. U 1994. godini nastavljen je rad na razradi i dopuni sovjetskog projekta. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Veoma obiman remont sekundarnog kola hladjenja reaktora privodi se kraju i bice okoncan u prvoj polovini 1995. a u skladu postojecim zakonskim propisima i sa preporukama MAAE. Kontrola goriva od strane inspektora MAAE obavljana je jedanput mesecno. Na reaktoru RA u 1994. godini radilo je prosecno 47 radnika, sto je dovoljan broj u uslovima remontnih i investicionih radova.

  19. Membrane reactor for water detritiation: a parametric study on operating parameters

    Energy Technology Data Exchange (ETDEWEB)

    Mascarade, J.; Liger, K.; Troulay, M.; Perrais, C. [CEA, DEN, DTN/STPA/LIPC, Centre de Cadarache, Saint-Paul-lez-Durance (France); Joulia, X.; Meyer, X.M. [Universite de Toulouse, INPT, UPS, Laboratoire de Genie Chimique, Toulouse (France); CNRS, Laboratoire de Genie Chimique, Toulouse (France)

    2015-03-15

    This paper presents the results of a parametric study done on a single stage finger-type packed-bed membrane reactor (PBMR) used for heavy water vapor de-deuteration. Parametric studies have been done on 3 operating parameters which are: the membrane temperature, the total feed flow rate and the feed composition through D{sub 2}O content variations. Thanks to mass spectrometer analysis of streams leaving the PBMR, speciation of deuterated species was achieved. Measurement of the amounts of each molecular component allowed the calculation of reaction quotient at the packed-bed outlet. While temperature variation mainly influences permeation efficiency, feed flow rate perturbation reveals dependence of conversion and permeation properties to contact time between catalyst and reacting mixture. The study shows that isotopic exchange reactions occurring on the catalyst particles surface are not thermodynamically balanced. Moreover, the variation of the heavy water content in the feed exhibits competition between permeation and conversion kinetics.

  20. Optimisation of deep burn incineration of reactor waste plutonium in a PBMR DPP-400 core

    International Nuclear Information System (INIS)

    Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik

    2014-01-01

    In this article an original set of coupled neutronics and thermo-hydraulic simulation results for the VSOP 99/05 diffusion code are presented for advanced fuel cycles for the incineration of weapons-grade plutonium, reactor-grade plutonium and reactor-grade plutonium with its associated Minor Actinides in the 400 MW th Pebble Bed Modular Reactor Demonstration Power Plant. These results are also compared to those of the standard 9.6 wt% enriched 9 g/fuel sphere U/Pu fuel cycle. The weapons-grade and reactor-grade plutonium fuel cycles produced good burn-ups. However, the addition of the Minor Actinides to the reactor-grade plutonium caused a large decrease in the burn-up and thus an unacceptable increase in the heavy metal (HM) content in the spent fuel, which was intended for direct disposal in a deep geological repository, without chemical reprocessing. All the plutonium fuel cycles failed the adopted safety limits used in the PBMR400 in that either the maximum fuel temperature of 1130 °C during normal operation, or the maximum power density of 4.5 kW/sphere was exceeded. All the plutonium fuel cycles also produced positive uniform temperature reactivity coefficients, i.e. the reactivity coefficient where the temperatures of the fuel and the graphite moderator in the fuel spheres were varied together. These unacceptable positive coefficients were experienced at low temperatures, typically below 700 °C. This was due to the influence of the thermal fission cross-section resonances of 239 Pu and 241 Pu. Weapons-grade plutonium produced the worst safety performance. The safety performance of the reactor-grade plutonium also deteriorated when the HM loading was reduced from 3 g/sphere to 2 g or 1 g

  1. Optimisation of deep burn incineration of reactor waste plutonium in a PBMR DPP-400 core

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School for Mechanical and Nuclear Engineering, North West University, PUK-Campus, Private Bag X6001, Internal Post Box 360, Potchefstroom 2520 (South Africa); Mulder, Eben J. [School for Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)

    2014-05-01

    In this article an original set of coupled neutronics and thermo-hydraulic simulation results for the VSOP 99/05 diffusion code are presented for advanced fuel cycles for the incineration of weapons-grade plutonium, reactor-grade plutonium and reactor-grade plutonium with its associated Minor Actinides in the 400 MW{sub th} Pebble Bed Modular Reactor Demonstration Power Plant. These results are also compared to those of the standard 9.6 wt% enriched 9 g/fuel sphere U/Pu fuel cycle. The weapons-grade and reactor-grade plutonium fuel cycles produced good burn-ups. However, the addition of the Minor Actinides to the reactor-grade plutonium caused a large decrease in the burn-up and thus an unacceptable increase in the heavy metal (HM) content in the spent fuel, which was intended for direct disposal in a deep geological repository, without chemical reprocessing. All the plutonium fuel cycles failed the adopted safety limits used in the PBMR400 in that either the maximum fuel temperature of 1130 °C during normal operation, or the maximum power density of 4.5 kW/sphere was exceeded. All the plutonium fuel cycles also produced positive uniform temperature reactivity coefficients, i.e. the reactivity coefficient where the temperatures of the fuel and the graphite moderator in the fuel spheres were varied together. These unacceptable positive coefficients were experienced at low temperatures, typically below 700 °C. This was due to the influence of the thermal fission cross-section resonances of {sup 239}Pu and {sup 241}Pu. Weapons-grade plutonium produced the worst safety performance. The safety performance of the reactor-grade plutonium also deteriorated when the HM loading was reduced from 3 g/sphere to 2 g or 1 g.

  2. Operational report, Advantages of gradual introducing of highly enriched fuel into the RA reactor core from economic aspect and users needs; Radni izvestaj, Prednost postupka parcijalnog uvodjenja visokoobogacenog goriva u reaktor RA sa aspekta ekonomicnosti i potreba korisnika

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-10-14

    The possibility of increasing the neutron flux in the RA reactor was considered for a number of years. The possibilities of reactor reconstruction are not realistic and they should be disregarded. The possibility that remains is to achieve higher neutron flux by improving the fueling scheme and above all by introducing highly enriched fuel into the reactor core. Decision to purchase highly enriched fuel was quicker due to the fact that the 2% enriched uranium fuel is not fabricated any more. There are two procedures for exchanging the fuel in the reactor core: a) removal of partially spent 2% enriched fuel and formation of the core with fresh highly enriched fuel; b) gradually introducing the new fuel into the existing RA reactor core according to a special transfer regime. This report includes some comparative analyses of these two procedures from both economic point of view and the needs of users, as well as some technical conditions. These results are in favour of gradual introducing of new fuel into the reactor core. relevant direct savings amount to 3 000 000 dinars. Some of the most important advantages cannot be estimated in this way. This report does not cover the safety analyses results which are presented in a series of other papers. [Serbo-Croat] Vec vise godina razmatra se mogucnost za povecanje neutronskog fluksa u reaktoru RA. Mogucnosti za rekonstrukciju reaktora RA u tom smislu su minimalne i realno ih treba odbaciti. Prema tome preostaje da se povecanje neutronskog fluksa postigne usavrsavanjem seme izmene goriva, a pre svega uvodjenjem goriva sa visokim stepenom obogacenja u reaktor RA. Donosenje odluke o nabavci visokoobogacenog goriva i njegovom uvodjenju u reaktor ubrzano je i cinjenicom da se staro 2% obogaceno uransko gorivo vise ne proizvodi. Postoje dva postupka za prevodjenje reaktora na ovo gorivo: a) Uklanjanjem poluistrosenog 2% obogacenog goriva iz reaktora i formiranjem jezgra iskljucivo od svezeg visokoobogacenog goriva, b

  3. Development of a neutronic model for the fuel of a high temperature gas reactor type PBMR

    International Nuclear Information System (INIS)

    Oropeza C, I.; Carmona H, R.; Francois L, J. L.

    2008-01-01

    In this work was developed the neutronic model of a fuel sphere of a nuclear reactor of gas of high temperature to modulate of bed of spheres (PBMR), using the Monte Carlo method with the MCNPx code. In order to be able to verify the fuel model constructed in this investigation, it is used a case of reference, based on an international exercise b enchmark . The benchmark report contains the results sent by different international participants for five phases with respect to the high temperature gas reactor (HTR), fed with uranium, plutonium and thorium. In particular, in first stage of benchmark an infinite adjustment of uranium compound fuel spheres is considered unique, with which our results were compared. This first stage considers two cases: cell calculations with spherical external frontier and cell calculations with cubic external frontier. The objective is to identify any increase in the uncertainty, related to the uranium fuel, that is associated with the plutonium and thorium fuels. In order to validate our results, the values of the neutron multiplication factor were taken in account, in cold and in the heat of the moment from the participants who sent their results obtained with Monte Carlo and deterministic calculations. The model of the fuel sphere developed in this work considers a regular distribution of 15000 Triso particles, in a cubic mesh centered within the sphere. For it was necessary to define the step firstly or p itch o f the cubic mesh. Generally, the results obtained by the participants of benchmark and those of this investigation present good agreement, nevertheless, appear some discrepancies, attributed to factors like different libraries of cross sections used, the nature of the solution: Monte Carlo or deterministic, and the difficulty of some participants to model the external frontier condition of reflection. (Author)

  4. Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Brian David [Los Alamos National Laboratory; Beddingfield, David H [Los Alamos National Laboratory; Durst, Philip [INL; Bean, Robert [INL

    2010-01-01

    The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

  5. A pulsed fast reactor; Un reacteur pulse a neutrons rapides; Impul'snyj reaktor na bystrykh nejtronakh; Reactor rapido pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, G. E.; Blokhintsev, D. I.; Blyumkina, Yu. A.; Bondarenko, I. I.; Deryagin, B. N.; Zajmovskij, A. S.; Zinov' ev, V. P.; Kazachkovskij, O. D.; Krasnoyarov, N. V.; Lejpunskij, A. I.; Malykh, V. A.; Nazarov, P. M.; Nikolaev, S. K.; Stavisskij, Yu. Ya.; Ukraintsev, F. I.; Frank, I. M.; Shapiro, F. Ji.; Yazvitskij, Yu. S. [Akademiya Nauk, Moscow, SSSR (Russian Federation)

    1962-03-15

    A pulsed fast reactor (IBR) has been operating at rated capacity since December 1960 in the Joint Institute for Nuclear Research. This reactor is used as a pulsed neutron source for physical experiments carried out by the time-of-flight method. It is used for total cross-section and intermediate neutron capture cross- section measurements, for studying the interaction between slow neutrons and solids and liquids, and for measuring neutron spectra produced in various media. The paper describes the basic structural features of the reactor and the results of the experiments for which it has been used. The reactor's operating system is based on recurrent pulses. Power pulses are produced when the mobile part of the reactor core moves swiftly through the stationary part of the core. The mobile part of the core is fastened to a rotating disc and travels at a speed of 230 m/s. The frequency of power pulses can be altered by means of an auxiliary mobile zone which has a range of 2.3-88 pulses per second. The mean power of the reactor is 1 kW, and the half-width of the power pulse in 36 {mu}s. The reactor is provided with a control and safety system which ensures automatic maintenance of mean power and swift shutdown in the event of any operational irregularity. It is fitted with a system of evacuated-neutron-flight tubes used in time-of-flight experiments. The main tube is 1000 m in length. In the start-up process and during physical experiments carried out on the reactor, the influence on reactivity of displacing the controls and the mobile parts of the core was studied ; the length of the pulse was measured under various operating conditions, and power pulse amplitude fluctuations were studied. Further measurements were made to establish the lifetime of prompt neutrons, the effective fraction of delayed neutrons, and coefficients of reactivity. (author) [French] L'Institut unifie de recherches nucleaires dispose d'un reacteur puise a neutrons rapides (IBR), qui

  6. Presentation summary, safety design aspects and U.S. licensing challenges of the pebble bed modular reactor

    International Nuclear Information System (INIS)

    Sproat, Ward; Slabber, Johan

    2001-01-01

    This presentation consists of three sections: An overview of the status of the PBMR project in South Africa, a review of the design features and philosophy being utilized to design the PBMR, and a summary of the key licensing issues that Exelon has identified in assessing the licensability of the PBMR for application in this country

  7. Studi Efisiensi Penyisihan COD dalam Lindi dengan Sistem Evapotranspirasi Menggunakan Tumbuhan Sente (Alocasia macrorrhiza dan Rumput Belulang (Eleusine indica

    Directory of Open Access Journals (Sweden)

    Badrus Zaman

    2017-11-01

    Full Text Available COD dalam lindi merupakan salah satu parameter yang secara umum berada pada konsentrasi yang tiggi sebagai salah satu hasil biodegradasi material organik dan anorganik dalam sampah di TPA. Sistem evapotranspirasi yang menggunakan tumbuhan lokal merupakan salah satu sistem yang menjanjikan. Penelitian dilakukan untuk mengetahui efisiensi penyisihan COD dalam lindi dengan reaktor evapotranspirasi secara kontinyu yang menggunakan tumbuhan Tumbuhan Sente (Alocasia macrorrhiza dan Rumput Belulang (Eleusine indica. Hasil uji menunjukkan efisiensi pada semua reaktor mulai sekitar hari ke 3 hingga hari ke 25 mengalami fluktuasi yang cenderung menurun (dari ± 75% menjadi ± 50%, tetapi hari selanjutnya cenderung meningkat. Pola tersebut dipengaruhi oleh peran media tanam, bakteri dalam media tanam, bakteri pada akar tumbuhan dan aktivitas metabolisme tumbuhan uji. Secara keseluruhan reaktor yang menggunakan Tumbuhan Sente (Alocasia macrorrhiza lebih fluktuatif dibandingkan denga menggunakan Rumput Belulang (Eleusine indica yang dipengaruhi pola pertumbuhan dan perkembangannya.

  8. RA Research nuclear reactor, Part 1, RA reactor operation and maintenance in 1993, with comparative review for the period 1991 - 1993, Annex 3; Projekat Istrazivacki nuklearni reaktor RA - 1 Deo Pogon i odrzavanje nuklearnog reaktora RA u 1993. godini, uz uporedni pregled za period 1991 - 1993. - prilog 3

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Mikic, N; Tanaskovic, M [Vinca Institute of Nuclear Sciences, Beograd (Serbia and Montenegro)

    1993-12-15

    RA reactor was not operated during 1993 because of the complete instrumentation exchange. Although it has been planned to exchange the complete instrumentation until the end of 1993, and to start reactor operation in the first half of 1993 this was not fulfilled because the instrumentation was not delivered until the end of 1993. Main activities during past seven years were related to construction of the emergency cooling system; repair and reconstruction of the system for handling the spent fuel and improvement of spent fuel storage conditions; exchange of the aged instrumentation. Other reactor components and systems, reactor core, primary coolant loop and gas circulation system are in good condition concerning future start-up. [Serbo-Croat] U 1993. godini reaktor nije bio u pogonu zbog zamene njegove celokupne instrumentacije. Iako je bilo planirano da se celokupna instrumentacija zameni do kraja 1993. te da reaktor pocne sa radom u prvoj polovini 1993. Ovo nije ispunjeno jer celokupna oprema nije isporucena ni do kraja 1993. godine. Osnovni zahvati koji su u proteklih sedma godina izvrseni, odnosili su se na izgradnju sistema za udesno hladjenje, rekonstrukciju sistema za rukovanje ozracenim gorivom i poboljsanje uslova za stokiranje ovog goriva, zamenu instrumentacije. Ostali sistemi reaktora, reaktorsko jezgro, primarno kolo hladjenja i sistem za cirkulaciju gasa su u dobrom stanju i mogu se nesmetano koristiti u buducem radu.

  9. The ESKOM pebble bed modular reactor

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1999-01-01

    An audit has been made of the design, construction, safety, economics and marketability of the ESKOM pebble bed modular reactor (PBMR). In this paper that audit is briefly summarized. The principal conclusions of the audit are as follows. The design is sound. It is a logical development of the designs proposed for other, modern, high-temperature gas-cooled reactors. More than 80% of the cost of constructing and commissioning a series of PBMRs would be spent in South Africa. The PBMR is much safer than existing nuclear power reactors and for many practical purposes it may be treated as a conventional chemical plant. The PBMR is economically competitive with thermal power stations. There is a substantial global market for the PBMR. (author)

  10. Generation 4 - nuclear reactors and an approach to secure public acceptance and access to energy for everyone

    International Nuclear Information System (INIS)

    Pahladsingh, R.

    2001-01-01

    The aim of this paper is to bring the Pebble Bed Modular Reactor (PBMR) and a few interesting Light Water Passive nuclear reactor designs under your attention. The PBMR is under further development in South Africa and Asia. The philosophy behind the PBMR concept has been to develop a nuclear reactor which is so safe that it could be called inherently safe. Its concept is so completely different, see figure 2, that it can easily pass strictest safety regulations. Consequently it is a good Generation IV candidate. Good promotion of the gas-turbine direct cycle PBMR design is a main task to the nuclear technology and industry and could be the challenge that the young generation needs to consider a career in nuclear technology. (authors)

  11. THERMAL-HYDRAULIC ANALYSIS OF SMR WITH NATURALLY CIRCULATING PRIMARY SYSTEM DURING LOSS OF FEED WATER ACCIDENT

    Directory of Open Access Journals (Sweden)

    Susyadi Susyadi

    2016-09-01

    ABSTRAK Reaktor daya kecil modular (SMR memiliki beberapa keunggulan dibanding reaktor daya besar konvensional. Dengan disain yang lebih sederhana dan terintegrasi, penerapan hukum alamiah untuk sistem keselamatannya dan biaya modal yang rendah, reaktor ini sangat cocok untuk dibangun di Indonesia. Salah satunya disain SMR yang sedang dikembangkan menerapkan gaya penggerak alami untuk sistim pendingin primernya. Dengan disain seperti itu, adalah sangat penting untuk memahami implikasinya terhadap aspek keselamatan pada seluruh kondisi operasi. Salah satu yang perlu diinvestigasi adalah kecelakaan kehilangan air umpan (LoFW. Pada studi ini, dilakukan analisis kinerja thermal hidrolik SMR yang menggunakan sistim pendinginan primer sirkulasi alam saat kecelakaan LoFW. Tujuannya adalah untuk menginvestigasi karakteristik aliran sistem primer saat kecelakaan LoFW dan untuk memastikan apakah aliran sirkulasi alam cukup untuk memindahkan panas dari teras guna menjaga kondisi tetap aman selama kecelakaan tersebut. Metoda yang digunakan adalah dengan merepresentasikan sistem reaktor ke dalam model-model generik program RELAP5 dan melakukan simulasi numerik. Hasil perhitungan menunjukkan bahwa setelah kejadian pemicu dan trip reaktor, pada sisi primer laju alirnya berfluktuasi secara signifikan dan temperatur pendinginnya menurun secara bertahap sedangkan  pada sisi sekunder kondisi uap berubah menjadi uap jenuh. Laju alir turun dari ~711 kg/detik menjadi ~263 kg/detik sebelum kembali naik lagi pada t=~46 detik. Saat laju alir di titik terendah, temperatur pusat bahan bakar dan fluida pendingin adalah sekitar  ~565 K dan  ~554 K, yang menujukkan bahwa temperatur bahan bakar masih jauh di bawah batas disain dan temperatur fluidanya juga berada di bawah titik saturasi. Keadaan ini menunjukkan bahwa saat transien kedua parameter utama termohidrolik reaktor tetap dalam kondisi yang dapat diterima sehingga dapat disimpulkan  bahwa saat  kecelakaan kehilangan air umpan, SMR

  12. RELAP5 SIMULATION FOR SEVERE ACCIDENT ANALYSIS OF RSG-GAS REACTOR

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2018-01-01

    SIMULASI RELAP5 UNTUK ANALISIS KECELAKAAN PARAH PADA REAKTOR RSG-GAS. Reaktor riset di dunia diketahui lebih aman dari pada reaktor daya karena desainnya yang lebih sederhana pada teras dan karakteristika operasinya. Namun demikian, potensi bahaya reaktor riset terhadap publik dan lingkungan tidak bisa diabaikan karena beberapa fitur tertentu. Oleh karena itu, level keselamatan reaktor riset harus jelas ditunjukkan dalam Laporan Analisis Keselamatan (LAK dalam bentuk analisis keselamatan yang dilakukan dengan berbagai macam pendekatan dan metode dan didukung dengan alat komputasi. Tujuan penelitian ini adalah untuk mensimulasikan beberapa kecelakaan parah pada reaktor RSG-GAS yang dapat menyebabkan kerusakan bahan bakar untuk memperkuat hasil analisis kecelakaan parah yang sudah ada dalam LAK. Simulation dilakukan dengan program perhitungan RELAP5/SCDAP/Mod3.4 yang memiliki kemampuan untuk memodelkan elemen bahan bakar tipe pelat di RSG-GAS. Tiga kejadian telah disimulasikan yaitu hilangnya aliran primer dan sekunder dengan kegagalan reaktor untuk dipadamkan, tersumbatnya beberapa kanal pendingin bahan bakar pada daya penuh, dan hilangnya aliran primer dan sekunder yang diikuti dengan tersumbatnya beberapa kanal pendingin bahan bakar setelah reaktor padam. Kejadian pertama akan membahayakan pelat bahan bakar dengan naiknya temperatur kelongsong hingga titik lelehnya yaitu 590 °C. Tersumbatnya satu atau beberapa kanal pada satu elemen bahan bakar menyebabkan konsekuensi yang berbeda pada pelat bahan bakar, dimana paling sedikit tersumbatnya 2 kanal akan merusak satu pelat bahan bakar, apalagi tersumbatnya satu elemen bahan bakar. Kombinasi antara hilangnya aliran pendingin primer dan sekunder yang diikuti dengan tersumbatnya satu kanal bahan bakar setelah reaktor dipadamkan menyebabkan naiknya temperatur kelongsong di bawah titik lelehnya yang berarti sirkulasi alam yang terbentuk dan daya yang terus turun cukup untuk mendinginkan elemen bahan bakar. Kata kunci

  13. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983; Istrazivacki nuklearni reaktor RA - Deo I - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1983. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Kozomara-Maic, S; Cupac, S; Raickovic, N; Radivojevic, J; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1983-12-15

    After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW. [Serbo-Croat] Pregledom nuklearnog goriva iz tehnoloskih kanala reaktora RA koji je izvrsen u periodu decembear 1982-feburuar 1983. godine nakon zaustavljanja reaktora po isteku novembarske kampanje 1982. godine, ustanovljeno je da ponovo dolazi do stvaranja taloga u obliku hidratisanih oksida aluminiuma na kosuljicama gorivnih elemenata. Nakon ponovnog pustanja u rad, reaktor je do novembra 1981. godine neprekidno bio u pogonu na snagama 1,8 - 2 MW. Jezgro je bilo formirano iskljucivo sa od gorivnih elemenata sa 80% obogacenim uran dioksidom. Utvrdjeno je da kosuljica gorivnog elementa nije korodirala, i da se bez nastval rada ne moze utvrditi uzrok pojave taloga na povrsini kosuljice. Da bi se mogucnost korozije aluminjumskih komponenti u primarnom kolu raktora svela na sto manju meru odluceno je da se vrsi neprekidno preciscavanje teske vode i da se istovremeno poveca protok teske vode radom dve pumpe, Komitet za sigurnost Instituta odobrio je ovakav nacin

  14. Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

    Science.gov (United States)

    Mkhabela, Peter Tshepo

    The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis

  15. Research nuclear reactor RA - Annual Report 1997; Istrazivacki nuklearni reaktor RA - Izvestaj za 1997. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Vinca Institute of Nuclear Sciences, Beograd (Serbia and Montenegro)

    1997-12-01

    RA reactor is not in operation since 1984, activities related to revitalisation of the RA reactor started in 1986. The planned actions related to renewal of the reactor components were finished except for the most important action, related to exchange of complete reactor instrumentation which was delayed. Only 80% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the sanctions imposed to our country. The existing RA reactor instrumentation was dismantled. Control and maintenance of the reactor components was done regularly and efficiently. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 42 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. The problem of financing the reactor activities and maintenance remains unsolved. Research reactor RA Annual report for year 1997 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Reaktor RA nije u pogonu od 1984, aktivnosti na revitalizaciji, rekostrukciji i modernizaciji reaktorski sistema zapocete su 1986. godine. Okoncan je niz zahvata na opremi postrojenja kojima ce se u narednom periodu omoguciti kontinualan i pouzdan rad ovog reaktora. Poslednji, i ujedno najveci zahvat, koji se odnosi na zamenu celokupne instrumentacije kasni zbog zastoja u isporuci opreme koja se izradjuje u Sovjetskom savezu. Do septembra 1991. godine isporuceno svega 80% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju usled sankcija uvedenih od strane organizacije Ujedinjenih nacija. Demontirana je postojeca instrumentacija. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Kontrola goriva od strane safeguard inspektora MAAE obavljana

  16. From field to factory-Taking advantage of shop manufacturing for the pebble bed modular reactor

    International Nuclear Information System (INIS)

    Wallace, Edward; Matzie, Regis; Heiderd, Roger; Maddalena, John

    2006-01-01

    The move of nuclear plant construction from the field to the factory for small, advanced pebble bed modular reactor (PBMR) designs has significant benefits compared to traditional light water reactor (LWR) field oriented designs. The use of modular factory construction techniques has a growing economic benefit over time through well-established process learning applications. This paper addresses the basic PBMR design objectives and commercialization model that drive this approach; provides a brief technical description of the PBMR design and layout with representative CAD views and discusses derived figures of merit highlighting the relative simplicity of PBMR compared to a modern LWR. The discussion emphasizes that more of PBMR can be built in the factory due to the simple design of a direct helium Brayton cycle compared to an indirect LWR steam cycle with its associated equipment. For the PBMR design there are fewer and less cumbersome auxiliary and safety systems with their attendant support requirements. Additionally, the labor force economic efficiency for nuclear projects is better in the factory than in the field, including consideration of labor costs and nuclear quality programs. Industrial learning is better in the factory because of the more controlled environment, mechanization optimization opportunities and because of the more stable labor force compared to the field. Supply chain benefits are more readily achievable with strategic contracts for module suppliers. Although building a nuclear power plant is not a typical high volume manufacturing process, for the PBMR-type of plant, with its high degree of standardization and relatively small, simplified design, the shift to factory work has a significant impact on overall project cost due to earlier identification and better coordination of parallel construction paths. This is in stark contrast to the construction of a large LWR in the past. Finally, the PBMR modular plant concept continues at the

  17. Efek Durasi Pencahayaan Pada Sistem HRAR Untuk Menurunkan Kandungan Minyak Solar Dalam Air Limbah

    Directory of Open Access Journals (Sweden)

    Dian Puspitasari

    2014-09-01

    Full Text Available Kandungan minyak di dalam air limbah industri perminyakan umumnya bersifat toksik terhadap mikroorganisme dan mengganggu proses pengolahan secara biologis. Sistem HRAR diperkirakan dapat mengatasi hambatan tersebut melalui proses fotosintesis untuk menghasilkan oksigen yang dibutuhkan mikroorganisme dalam mendegradasi senyawa hidrokarbon. Penelitian ini bertujuan mengkaji pengaruh perpanjangan waktu pencahayaan pada kemampuan HRAR dalam menurunkan kandungan minyak di dalam limbah. Variabel yang digunakan pada penelitian ini adalah variasi durasi pencahayaan dan variasi penambahan volume minyak solar yang ditambahkan ke dalam reaktor. Variasi durasi pencahayaan yang digunakan adalah pencahayaan selama 12 jam dan pencahayaan selama 24 jam. Sedangkan penambahan volume minyak solar ke dalam masing-masing reaktor adalah sebesar 346 ppm, 519 ppm dan 692 ppm. Hasil yang didapatkan dari penelitian ini adalah durasi pencahayaan selama 12 jam memiliki efek yang lebih baik terhadap penurunan konsentrasi minyak dibandingkan pencahayaan selama 24 jam. Hal ini dapat terlihat dari baiknya pertumbuhan alga dan bakteri di dalam reaktor serta tingginya penurunan konsentrasi minyak solar di dalamnya. Penurunan konsentrasi minyak solar terbaik terdapat pada reaktor dengan penambahan minyak solar sebesar 346 ppm. Pada reaktor dengan durasi pencahayaan selama 12 jam terjadi penurunan konsentrasi minyak sebesar 78,4%. Sedangkan penurunan kandungan minyak solar pada reaktor dengan durasi pencahayaan selama 24 jam adalah sebesar 73,9%.

  18. The Text of the Agreement of 22 July 1977 between Argentina and the Agency for the Application of Safeguards in Connection with a Contract Concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union Gmbh Hanau (Federal Republic of Germany) for Co-Operation in the Field of Fabrication of Fuel Elements for Peaceful Nuclear Activities

    International Nuclear Information System (INIS)

    1977-01-01

    The text of the Agreement of 22 July 1977 between Argentina and the Agency for the application of safeguards in connection with the Contract of 13 August 1976 concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union GmbH (Federal Republic of Germany) for co-operation in the field of fabrication of fuel elements for peaceful nuclear activities is reproduced in this document for the information of all Members. The Agreement entered into force, pursuant to Section 26, on 22 July 1977.

  19. The Text of the Agreement of 22 July 1977 between Argentina and the Agency for the Application of Safeguards in Connection with a Contract Concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union Gmbh Hanau (Federal Republic of Germany) for Co-Operation in the Field of Fabrication of Fuel Elements for Peaceful Nuclear Activities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-11-30

    The text of the Agreement of 22 July 1977 between Argentina and the Agency for the application of safeguards in connection with the Contract of 13 August 1976 concluded between the Comision Nacional de Energia Atomica (Argentina) and the Reaktor Brennelement Union GmbH (Federal Republic of Germany) for co-operation in the field of fabrication of fuel elements for peaceful nuclear activities is reproduced in this document for the information of all Members. The Agreement entered into force, pursuant to Section 26, on 22 July 1977.

  20. Fuel analysis of a PBMR; Analisis de combustible de un PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Bastida O, G. E.; Francois L, J. L., E-mail: gbo729@yahoo.com.mx [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

    2015-09-15

    In this paper a neutronic analysis of fuel for a Pebble Bed Modular Reactor is presented, based on their composition and geometric distribution, having as main objective the use and utilization of thorium for the production of fuel for the operation of this reactor. For the study of these characteristics is necessary to use a code capable of carry out a reliable calculation of the main parameters of the fuel. Using the Monte Carlo method is suitable for simulating the neutron transport in the reactor core, which is the basis of Serpent code, with which the calculations for the analysis will be made. The results show the desirability of the use of thorium, since presents good conversion levels of fertile material to fissile, to produce U{sup 233} by neutron capture, taking as a very important factor the distribution of materials in the core, which in this work had better results based on the neutron multiplication effective factor, formed by three right circular cylinders circumscribed, making that the core has three areas constituted by a mixture of plutonium oxide in the central and external areas, and thorium oxide in the intermediate area. (Author)

  1. Importance of the multi-modules study in PSA

    International Nuclear Information System (INIS)

    Gonzalez R, V. J.; Nelson E, P. F.

    2015-09-01

    The current approach that has taken the Probabilistic Safety Analysis (PSA) consists of doing all the APS analysis including the existence of multi-units in the nuclear power plants (NPP), this new approach seeks to analyze the risk of site, evaluating all reactors together. The main reasons for this trend are: the accident occurred on March 2011 in Fukushima Daiichi in Japan, with serious consequences in more than one reactor of the NPP and the current planning and construction of new Small Modular Reactors, which host more than one module on the same NPP and are connected to a single control room. This study analyzes how to model the risk of a multi-module NPP. In 2013, the ASME/ANS standard for advanced reactors that are not light-water reactors was published, in which the requirements to realize a PSA including multi-units or modules are shown; however, does not describe the methodology to do that. This article presents a methodology to calculate the risk of the site in a PBMR plant with two modules. This methodology consists of two models of trees of different events, one that evaluates to a single PBMR module and another that evaluates the two modules together. Both models are responsible to show their differences and compare results to finally demonstrate the need for new methodologies for risk analysis site in multi-modules and units. (Author)

  2. PEMODELAN KOLIMATOR DI RADIAL BEAM PORT REAKTOR KARTINI UNTUK BORON NEUTRON CAPTURE THERAPY

    Directory of Open Access Journals (Sweden)

    Bemby Yulio Vallenry

    2015-03-01

    Full Text Available Salah satu metode terapi kanker adalah Boron Neutron Capture Therapy (BNCT. BNCT memanfaatkan tangkapan neutron oleh 10B yang terendapkan pada sel kanker. Keunggulan BNCT dibandingkan dengan terapi radiasi lainnya adalah tingkat selektivitas yang tinggi karena tingkatannya adalah sel. Pada penelitian ini dilakukan pemodelan kolimator di radial beamport reaktor Kartini sebagai dasar pemilihan material dan manufature kolimator sebagai sumber neutron untuk BNCT. Pemodelan ini dilakukan dengan simulasi menggunakan perangkat lunak Monte Carlo N-Particle versi 5 (MCNP 5. MCNP 5 adalah suatu paket program untuk memodelkan sekaligus menghitung masalah transpor partikel dengan mengikuti sejarah hidup neutron semenjak lahir, bertranspor pada bahan hingga akhirnya hilang karena mengalami reaksi penyerapan atau keluar dari sistem. Pemodelan ini menggunakan variasi material dan ukurannya agar menghasilkan nilai dari tiap parameter-parameter yang sesuai dengan rekomendasi I International Atomic Energy Agency (IAEA untuk BNCT, yaitu fluks neutron epitermal (Фepi > 9 n.cm-2.s-1, rasio antara laju dosis neutron cepat dan fluks neutron epitermal (Ḋf/Фepi 0,7. Berdasarkan hasil optimasi dari pemodelan ini, material dan ukuran penyusun kolimator yang didapatkan yaitu 0,75 cm Ni sebagai dinding kolimator, 22 cm Al sebagai moderator dan 4,5 cm Bi sebagai perisai gamma. Keluaran berkas radiasi yang dihasilkan dari pemodelan kolimator radial beamport yaitu Фepi = 5,25 x 106 n.cm-2s-1, Ḋf/Фepi =1,17 x 10-13 Gy.cm2.n-1, Ḋγ/Фepi = 1,70 x 10-12 Gy.cm2.n-1, Фth/Фepi = 1,51 dan J/Фepi = 0,731. Berdasarkan penelitian ini, hasil optimasi 5 parameter sebagai persyaratan kolimator untuk BNCT yang keluar dari radial beam port tidak sepenuhnya memenuhi kriteria yang direkomendasikan oleh IAEA sehingga perlu dilakukan penelitian lebih lanjut agar tercapainya persyaratan IAEA. Kata kunci: BNCT, radial beamport, MCNP 5, kolimator   One of the cancer therapy methods is

  3. Research nuclear reactor RA - Annual Report 1995, with comparative review for period 1991-1995; Istrazivacki nuklearni reaktor RA - Izvestaj za 1995. godinu, uz uporedni pregled za period 1991-1995

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1995-12-01

    firmom Atomenergoexport. Prema tom ugovoru trebalo je da celokupna nova instrumentacija za reaktor RA bude isporucena Institutu u Vinci do kraja 1990. godine, ali je do septembra 1991. godine isporuceno svega 56% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju izrecena od strane ove organizacije Ujedinjenih nacija. Demontirana je postojeca instrumentacija u maksimalno mogucem obimu, kako bi se zadrzala neka osnovna merenja neophodna i u uslovima kada reaktor nije u pogonu. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Kontrola goriva od strane inspektora MAAE obavljana je jedanput mesecno. U ovom izvestajnom periodu na reaktoru RA bilo je zaposleno prosecno 47 radnika, sto je dovoljan broj u uslovima remontnih i investicionih radova. Nedostatak finansijskih sredstava za odrzavanje reaktora RA je neresen problem i u ovom periodu. Ovaj izvestaj podeljen je u dve celine: pogon i odrzavanje reaktora i zastita od zracenja na reaktoru RA.

  4. KARAKTERISTIK REAKTIVITAS TERAS KERJA RSG-GAS SELAMA 30 TAHUN BEROPERASI

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2017-06-01

    Full Text Available RSG-GAS, mulai dari komisioning, operasi teras kerja hingga kini telah 30 tahun beroperasi sehingga perlu dilakukan evaluasi keselamatan parameter neutroniknya. Untuk tujuan keselamatan telah dilakukan berbagai aktivitas penelitian, baik yang berhubungan dengan operasi, keselamatan, maupun dalam rangka penggunaan reaktor. Analisis dan pengelolaan besaran reaktivitas yang menunjang keselamatan operasi reaktor sangat penting dilakukan karena besaran ini mempengaruhi desain, kendali dan jadual operasi reaktor. Besaran tersebut dapat ditentukan melalui pengukuran reaktivitas batang kendali dan eksperimen pemuatan bahan bakar di dalam teras. Pengukuran reaktivitas batang kendali yang dilakukan pada setiap awal siklus teras (dengan kondisi teras dingin dan bersih, bebas pengaruh xenon, menghasilkan nilai reaktivitas batang kendali yang dapat digunakan untuk menentukan nilai reaktivitas lainnya seperti reaktivitas lebih, reaktivitas padam dan reaktivitas total. Pengelolaan reaktivitas teras telah dilakukan dengan baik selama 30 tahun dalam rangka mendukung operasi reaktor untuk keperluan penelitian dan iradiasi target.

  5. Research nuclear reactor RA - Annual Report 1990 with the comparative evaluation for the period 1986-1990; Istrazivacki nuklearni reaktor RA - Izvestaj za 1990. godinu uz uporedni pregled za period 1986 - 1990

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1990-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1990, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1990. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3705/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjjring RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  6. Integrated design approach of the pebble bed modular using models

    International Nuclear Information System (INIS)

    Venter, P.J.

    2005-01-01

    The Pebble Bed Modular Reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, showing the iterative design process that is used in the development of the reactor at PBMR. (author)

  7. External irradiation of the personnel operating the reactor RA at Vinca in the period 1963-1966; Spoljasnje ozracivanje osoblja koje opsluzuje reaktor RA u Vinci u periodu 1963-1966

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M M; Minincic, Z [Institut za nuklearne nauke Vinca, Belgrade (Yugoslavia)

    1967-07-01

    The objective of this paper was the analysis of external radiation exposure of the personnel employed at the RA reactor in Vinca during past four years. During 1963 reactor was not operated because of the general repair and maintenance, and it was operated during 1964, 1965 and 1966. The internal irradiation and the estimation of the beta doses on the skin were not analysed in this paper, they should be treated by separate analyses. The evaluation of external irradiation covered only gamma radiation since yhe neutron doses were negligible. The analysis was based on the the irradiation dose data obtained by film and personal dosemeters. Predmet ovog rada je analiza spoljasnjeg ozracivanja osoblja koje je radilo na reaktoru RA u Vinci u poslednje cetiri godine, u toku kojih je reaktor bio u opstem remontu 1963, odnosno u normalnom radu 1964, 1965 i 1966. U ovom radu nismo se upustali u analizu unutrasnjeg ozracivanja, kao ni procenu doza beta zracenja na kozi, sto treba da bude predmet posebnih analiza. Pri proceni spoljasnjeg ozracivanja operisano je samo sa dozama gama zracenja, jer je ozracivanje neutronima zanemarljivo. Celokupna analiza izvrsena je na osnovu podataka o dozama ozracivanja do kojih se doslo ocitavanjem film o penkalo dozimetara.

  8. High-Q plasmonic bottle microresonator

    Science.gov (United States)

    Mohd Nasir, M. Narizee; Ding, Ming; Murugan, G. Senthil; Zervas, Michalis N.

    2014-03-01

    In this paper, we demonstrate a hybrid plasmonic bottle microresonator (PBMR) which supports whispering gallery modes (WGMs) along with surface plasmon waves (SPWs) for high performance optical sensor applications. The BMR was fabricated through "soften-and-compress" technique with a thin gold layer deposited on top of the resonator. A polarization-resolved measurement was set-up in order to fully characterize the fabricated PBMR. Initially, the uncoated BMR with waist diameter of 181 μm, stem diameter of 125 μm and length of 400 μm was fabricated and then gold film was deposited on the surface. Due to surface curvature, the gold film covering half of the BMR had a characteristic meniscus shape and maximum thickness of 30 nm. The meniscus provides appropriately tapered edges which facilitate the adiabatic transformation of BMR WGMs to SPWs and vice versa. This results in low transition losses, which combined with partially-metal-coated resonator, can result in high hybrid-PBMR Q's. The transmission spectra of the hybrid PBMR are dramatically different to the original uncoated BMR. Under TE(TM) excitation, the PBMR showed composite resonances with Q of ~2100(850) and almost identical ~ 3 nm FSR. We have accurately fitted the observed transmission resonances with Lorentzian-shaped curves and showed that the TE and TM excitations are actually composite resonances comprise of two and three partially overlapping resonances with Q's in excess of 2900 and 2500, respectively. To the best of our knowledge these are the highest Qs observed in plasmonic microcavities.

  9. ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

    Directory of Open Access Journals (Sweden)

    Setiyanto Setiyanto

    2016-10-01

    Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

  10. Study for Safeguards Challenges to the Most Probably First Indonesian Future Power Plant of the Pebble Bed Modular Reactor

    International Nuclear Information System (INIS)

    Susilowati, E.

    2015-01-01

    In the near future Indonesia, the fourth most populous country, plans to build a small size power plant most probably a Pebble Bed Modular Reactor PBMR. This first nuclear power plant (NPP) is aimed to provide clear picture to the society in regard to performance and safety of nuclear power plant operation. Selection to the PBMR based on several factor including the combination of small size of the reactor and type of fuel allowing the use of passive safety systems, resulting in essential advantages in nuclear plant design and less dependence on plant operators for safety. In the light of safeguards perspective this typical reactor is also quite difference with previous light water reactor (LWR) design. From the fact that there are a small size large number of elements present in the reactor produced without individual serial numbers combine to on-line refueling same as the CANDU reactor, enforcing a new challenge to safeguards approach for this typical reactor. This paper discusses a bunch of safeguards measures have to be prepared by facility operator to support successfully international nuclear material and facility verification including elements of design relevant to safeguards need to be accomplished in consultation to the regulatory body, supplier or designer and the Agency/IAEA such as nuclear material balance area and key measurement point; possible diversion scenarios and safeguards strategy; and design features relevant to the IAEA equipment have to be installed at the reactor facility. It is deemed that result of discussion will alleviate and support the Agency approaching safeguards measure that may be applied to the purpose Indonesian first power plant of PBMR construction and operation. (author)

  11. Integrated design approach of the pebble BeD modular reactor using models

    International Nuclear Information System (INIS)

    Venter, Pieter J.; Mitchell, Mark N.

    2007-01-01

    The pebble bed modular reactor (PBMR) is the first pebble bed reactor that will be utilised in a high temperature direct Brayton cycle configuration. This implies that there are a number of unique features in the PBMR that extend from the German experience base. One of the challenges in the design of the PBMR is developing an understanding of the expected behaviour of the reactor through analyses and simulations and managing the integrated design process between the designers, the physicists and the analysts. This integrated design process is managed through model-based development work. Three-dimensional CAD models are constructed of the components and parts in the reactor. From the CAD models, CFD models, neutronic models, shielding models, FEM models and other thermodynamic models are derived. These models range from very simple models to extremely detailed and complex models. The models are used in legacy software as well as commercial off-the-shelf software. The different models are also used in code-to-code comparisons to verify the results. This paper will briefly discuss the different models and the interaction between the models, and how the models are used in the iterative design process that is used in the development of the reactor at PBMR

  12. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  13. Phenomenological modeling and study of a catalytic membrane reactor for water detritiation

    International Nuclear Information System (INIS)

    Mascarade, Jeremy

    2015-01-01

    Tritium is produced in light and heavy water reactor fuel by ternary fission or neutron activation. This by-product is used as fuel in fusion fuel reactors such as JET in Culham or ITER in Cadarache (France). The growing interest of this research area will make the tritium fluxes increase; it is then worth addressing the question of its future whether it will be used or flushed out from liquid and gaseous effluents or waste. This thesis studies the recovery of tritium as fuel for fusion machines by means of packed bed membrane reactor (PBMR). Such a reactor combines catalytic conversion of tritiated water thanks to isotope exchange with hydrogen according to the reversible reaction Q 2 O+H 2 ↔H 2 O+Q 2 (Q=H,D or T) and selective permeation of Q 2 through Pd-based membrane. In fact, palladium has the ability to bond with hydrogen isotopes, creating a selective permeation barrier. In the PBMR, thanks to the reaction products withdrawal, these permeation fluxes drive the heavy water conversion rate, to higher values than those reached in conventional fixed bed reactors (Le Chatelier's law). In order to study PBMRs, the CEA has built a test bench, using deuterium instead of tritium, allowing the analysis of their conversion and separation performances at the laboratory scale. An in-house method has been developed to determine simultaneously hydrogen and water isotopologues content by mass spectrometer analysis. It was experimentally shown that the activity of Ni-based catalyst used in this study was sufficient to allow the isotope exchange reactions to reach their thermodynamic equilibrium in a very short time. In addition, hydrogen permeation flux was shown to follow a Richardson's law. Sensitivity studies performed on the PBMR's main operating parameters revealed that its global performance (i.e. de-deuteration factor) increases with the temperature, the transmembrane pressure difference, the sweep gas flow rate and the residence time in the catalyst

  14. The Pebble Bed Modular Reactor: An obituary

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Steve, E-mail: stephen.thomas@gre.ac.u [Public Services International Research Unit (PSIRU), Business School, University of Greenwich, 30 Park Row, London SE10 9LS (United Kingdom)

    2011-05-15

    The High Temperature Gas-cooled Reactor (HTGR) has exerted a peculiar attraction over nuclear engineers. Despite many unsuccessful attempts over half a century to develop it as a commercial power reactor, there is still a strong belief amongst many nuclear advocates that a highly successful HTGR technology will emerge. The most recent attempt to commercialize an HTGR design, the Pebble Bed Modular Reactor (PBMR), was abandoned in 2010 after 12 years of effort and the expenditure of a large amount of South African public money. This article reviews this latest attempt to commercialize an HTGR design and attempts to identify which issues have led to its failure and what lessons can be learnt from this experience. It concludes that any further attempts to develop HTGRs using Pebble Bed technology should only be undertaken if there is a clear understanding of why earlier attempts have failed and a high level of confidence that earlier problems have been overcome. It argues that the PBMR project has exposed serious weaknesses in accountability mechanisms for the expenditure of South African public money. - Research highlights: {yields} In this study we examine the reasons behind the failure of the South African PBMR programme. {yields} The study reviews the technical issues that have arisen and lessons for future reactor developments. {yields} The study also identifies weaknesses in the accountability mechanisms for public spending.

  15. Importance of the multi-modules study in PSA; Importancia del estudio de multi-modulos en APS

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez R, V. J.; Nelson E, P. F., E-mail: judith_gonzalez_rodriguez@outlook.es [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2015-09-15

    The current approach that has taken the Probabilistic Safety Analysis (PSA) consists of doing all the APS analysis including the existence of multi-units in the nuclear power plants (NPP), this new approach seeks to analyze the risk of site, evaluating all reactors together. The main reasons for this trend are: the accident occurred on March 2011 in Fukushima Daiichi in Japan, with serious consequences in more than one reactor of the NPP and the current planning and construction of new Small Modular Reactors, which host more than one module on the same NPP and are connected to a single control room. This study analyzes how to model the risk of a multi-module NPP. In 2013, the ASME/ANS standard for advanced reactors that are not light-water reactors was published, in which the requirements to realize a PSA including multi-units or modules are shown; however, does not describe the methodology to do that. This article presents a methodology to calculate the risk of the site in a PBMR plant with two modules. This methodology consists of two models of trees of different events, one that evaluates to a single PBMR module and another that evaluates the two modules together. Both models are responsible to show their differences and compare results to finally demonstrate the need for new methodologies for risk analysis site in multi-modules and units. (Author)

  16. Comparative studies in farther-reaching waste water cleaning in different reactor systems; Vergleichende Untersuchungen zur weitergehenden Abwasserreinigung in unterschiedlichen Reaktorsystemen

    Energy Technology Data Exchange (ETDEWEB)

    Dockhorn, T. [Technische Univ. Braunschweig (Germany). Inst. fuer Siedlungswasserwirtschaft

    1999-07-01

    Three semi-technical pilot plants (completely mixed reactor, cascade, SBR) were operated in parallel under equal starting conditions. The influence of the type of reactor on the processes COD elimination, nitrification, denitrification and biological P elimination under operating conditions was studied. (orig.) [German] Es wurden drei halbtechnische Versuchsanlagen (volldurchmischter Reaktor, Kaskade, SBR) unter gleichen Ausgangsbedingungen parallel betrieben. Hierbei wurde der Einfluss des Reaktortyps auf die Prozesse CSB-Elimination, Nitrifikation, Denitrifikation sowie biologische P-Elimination unter Betriebsbedingungen untersucht. (orig.)

  17. EFEK DENSITAS BAHAN BAKAR TERHADAP PARAMETER KOEFISIEN REAKTIVITAS TERAS RRI

    Directory of Open Access Journals (Sweden)

    Rokhmadi Rokhmadi

    2015-03-01

    Full Text Available Manfaat yang luas penggunaan reaktor riset membuat banyak negara membangun reaktor riset baru. Kecenderungan saat ini adalah tipe reaktor serbaguna (MPR dengan teras yang kompak untuk mendapatkan fluks neutron yang tinggi dengan daya yang relatif rendah. Reaktor riset yang ada di Indonesia usianya sudah tua semuanya. Oleh karena itu diperlukan desain reaktor riset baru sebagai alternatif, disebut reaktor riset inovatif (RRI, kelak pengganti reaktor riset yang sudah ada. Tujuan dari riset ini untuk melengkapi data desain RRI sebagai salah satu persyaratan untuk perizinan desain. Perhitungan dilakukan untuk memperoleh nilai koefisien reaktivitas teras RRI dengan konfigurasi teras setimbang yang optimal dengan konfigurasi teras 5×5 dan daya 20 MW, memiliki panjang operasi satu siklus lebih dari 40 hari. Perhitungan koefisien reaktivitas teras RRI dilakukan untuk bahan bakar baru U-9Mo-Al dengan kerapatan bervariasi. Perhitungan dilakukan dengan paket program WIMSD-5B dan BATAN-FUEL. Hasil pehitungan digunakan untuk melengkapi data desain konseptual teras yang menunjukkan bahwa teras setimbang reaktor RRI dengan konfigurasi 5×5, tingkat muat 235U sebesar 450 g, 550 g dan 700 g memiliki nilai koefisien reaktivitas temperatur bahan bakar, temperatur moderator, densitas moderator dan void semuanya negatif dan nilainya sangat bervariasi. Hal ini sudah memenuhi kriteria keselamatan desain konseptual teras RRI. Kata kunci: desain konseptual, bahan bakar uranium-molibdenum, koefisien reaktivitas, WIMS, BATANFUEL   The multipurpose of research reactor utilization make many countries build the new reserach reactor. Trend of this reactor for this moment is multipurpose reactor type with a compact core to get high neutron flux at the low or medium level of power. The research reactor in Indonesia right now is already 25 year old. Therefor, it is needed to design a new research reactor as a alternative called it innovative research reactor (IRR and then as

  18. South Africa's nuclear model: A small and innovative reactor is seen as the model for new electricity plants. The project is nearing the starting blocks

    International Nuclear Information System (INIS)

    Ferreira, Tom

    2004-01-01

    Although nuclear power generation has by far the best safety and environmental record of any technology in general use, it has for many years been unable to make any meaningful inroads into the wall of negative perceptions that have arisen against it. But sentiments are changing rapidly on a global scale. The flare-up of oil prices is a sobering reminder of the volatility in the energy market, the exhaustibility of fossil fuels and the urgent need for stable, reliable, non-polluting sources of electrical power that are indispensable to a modern industrial economy. Today, new types of nuclear plants are prized, and South Africa is moving ahead. The State energy provider, Eskom, is internationally regarded as the leader in the field of the Pebble Bed Modular Reactor (PBMR) technology, a 'new generation' nuclear power plant. A decision on the PBMR project's future is on the near horizon. Should approvals be received in the coming months to proceed to the project's next phase, construction of the PBMR demonstration plant will start in 2006, in which case the reactor will start in 2010 and handed over to the client, Eskom, in 2011. Eskom has conditionally undertaken to purchase the first commercial units. Pebble bed reactors are small, about one-sixth the size of most current nuclear plants. Multiple PBMRs can share a common control center and occupy an area of no more than three football fields. More specifically, the PBMR is a helium-cooled, graphite moderated high temperature reactor (HTR). The concept is based on experience in the UK, United States and particularly Germany where prototype reactors were operated successfully between the late 1960s and 1980s. Although it is not the only high-temperature, gas-cooled nuclear reactor being developed in the world, the South African project is internationally regarded as a front-runner. The South African PBMR includes unique and patented technological innovations which make it particularly competitive. The Chief Executive

  19. The influence of annealing temperature on the strength of TRISO coated particles

    International Nuclear Information System (INIS)

    Rooyen, I.J. van; Neethling, J.H.; Rooyen, P.M. van

    2010-01-01

    The integrity of the Pebble Bed Modular Reactor (PBMR) fuel, and specifically the SiC layer system of the Tristructural Isotropic (TRISO) coated particle (CP), namely inner pyrolytic carbon, silicon carbide and outer pyrolytic carbon (I-PyC-SiC-O-PyC), determines the containment of fission products. The PBMR fuel consists of TRISO coated particles (CPs) embedded in a graphite matrix. One of the characterization techniques investigated by PBMR is the determination of strength of CPs. It is a well known metallurgical fact that temperature, amongst many other parameters, may influence the strength of a material. A recently developed method for measuring the strength of the TRISO coated particles was used and is briefly described in this article. The advantages of this method are demonstrated by the comparison of strength measurements of five experimental PBMR CP batches as a function of annealing temperature. Significant modification of strength after annealing was measured with increased temperature within the range 1000-2100 o C. The interesting feature of decreasing standard deviation of the strength with increasing temperature will also be discussed with a possible explanation. A significant difference in coated particle strength is also demonstrated for two CP batches with layer thickness on the extremities of the SiC layer thickness specification. The effect of long duration annealing on these strength values will also be demonstrated by comparing results from 1 h to 100 h annealing periods of coated particles at a temperature of 1600 o C.

  20. External irradiation of the personnel operating the reactor RA at Vinca in the period 1963-1966; Spoljasnje ozracivanje osoblja koje opsluzuje reaktor RA u Vinci u periodu od 1963. do 1966. godine

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M M; Minincic, Z [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1968-06-15

    The paper first gives a survey of the characteristic operations performed on the Vinca reactor RA in which most of the personnel become irradiated. Following is a schematic diagram of the irradiations in the period 1963-1966 in which the reactor was in continual operation. The surveys are given for each month and each year separately, while the irradiated personnel are grouped in several characteristic irradiation dose ranges. In this analysis special emphasis is given to a survey of irradiation of the personnel classified according to their profession, i.e. their post. This kind of analysis is indispensable in planning work, proper disposition of the personnel and undertaking special protective measures for reducing the irradiations (author) [Serbo-Croat] U radu je najpre dat pregled mesta na reaktoru RA, sa topografijom zracenja, na kojima radno osoblje pretezno biva ozracivano. Zatim su dati tabelarno i graficki pregledi ozracivanja za protekli period u kome je reaktor neprekidno radio, tj. od 1963. do 1966. god. Pregledi su dati po mesecima za svaku godinu i to tako sto su ozracivana lica razvrstavana u nekoliko karakteristicnih intervala doza ozracivanja. Posebno mesto u ovoj analizi zauzima pregled ozracivanja osoblja razvrstanog po strukama, odnosno radnim mestima. Ovakva analiza je nuzna za planiranje poslova, pravilan raspored radnog osoblja kao i preduzimanje posebnih zastitnih mera u cilju smanjivanja ozracivanja (author)

  1. FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE

    Directory of Open Access Journals (Sweden)

    Entin Hartini

    2016-06-01

    Full Text Available ABSTRACT FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE. The reactor pressure vessel (RPV is a pressure boundary in the PWR type reactor which serves to confine radioactive material during chain reaction process. The integrity of the RPV must be guaranteed either  in a normal operation or accident conditions. In analyzing the integrity of RPV, especially related to the crack behavior which can introduce break to the reactor pressure vessel, a fracture mechanic approach should be taken for this assessment. The uncertainty of input used in the assessment, such as mechanical properties and physical environment, becomes a reason that the assessment is not sufficient if it is perfomed only by deterministic approach. Therefore, the uncertainty approach should be applied. The aim of this study is to analize the uncertainty of fracture mechanics calculations in evaluating the reliability of PWR`s reactor pressure vessel. Random character of input quantity was generated using probabilistic principles and theories. Fracture mechanics analysis is solved by Finite Element Method (FEM with  MSC MARC software, while uncertainty input analysis is done based on probability density function with Latin Hypercube Sampling (LHS using python script. The output of MSC MARC is a J-integral value, which is converted into stress intensity factor for evaluating the reliability of RPV’s 2D. From the result of the calculation, it can be concluded that the SIF from  probabilistic method, reached the limit value of  fracture toughness earlier than SIF from  deterministic method.  The SIF generated by the probabilistic method is 105.240 MPa m0.5. Meanwhile, the SIF generated by deterministic method is 100.876 MPa m0.5. Keywords: Uncertainty analysis, fracture mechanics, LHS, FEM, reactor pressure vessels   ABSTRAK ANALISIS KETIDAKPASTIAN FRACTURE MECHANIC PADA EVALUASI KEANDALAN

  2. Fuel analysis of a PBMR

    International Nuclear Information System (INIS)

    Bastida O, G. E.; Francois L, J. L.

    2015-09-01

    In this paper a neutronic analysis of fuel for a Pebble Bed Modular Reactor is presented, based on their composition and geometric distribution, having as main objective the use and utilization of thorium for the production of fuel for the operation of this reactor. For the study of these characteristics is necessary to use a code capable of carry out a reliable calculation of the main parameters of the fuel. Using the Monte Carlo method is suitable for simulating the neutron transport in the reactor core, which is the basis of Serpent code, with which the calculations for the analysis will be made. The results show the desirability of the use of thorium, since presents good conversion levels of fertile material to fissile, to produce U 233 by neutron capture, taking as a very important factor the distribution of materials in the core, which in this work had better results based on the neutron multiplication effective factor, formed by three right circular cylinders circumscribed, making that the core has three areas constituted by a mixture of plutonium oxide in the central and external areas, and thorium oxide in the intermediate area. (Author)

  3. Perancangan dan Simulasi MRAC untui Proses Pengendalian Temperatur pada Continuous Stirred Tank Reactor (CSTR

    Directory of Open Access Journals (Sweden)

    Amelia Sylvia

    2014-03-01

    Full Text Available Temperatur merupakan salah satu variabel proses dasar yang dikendalikan untuk menjaga suhu cairan di dalam reaktor. Model Reference Adaptive Controller (MRAC dengan MIT rule dipilih untuk mencapai spesifikasi respon yang diinginkan pada CSTR. Beban yang bervariasi berupa debit aliran likuid yang masuk ke dalam reaktor dapat menyebabkan perubahan parameter yang mempengaruhi perubahan temperatur output produk pada CSTR. Sebuah simulasi dilakukan dengan menggunakan MATLAB dan hasilnya dianalisa. Respon plant dapat melakukan adaptasi parameter – parameter kontrolernya cukup baik pada nilai gain adaptasi dengan rentang 0.00000010000 sampai 0.00000000001. Waktu yang dibutuhkan untuk mengatasi beban yang bervariasi berupa debit aliran yang masuk ke dalam reaktor dengan nilai yang maksimal (1.5 m^3/min menghasilkan respon plant lebih cepat 42 detik dari pada debit aliran masuk dengan nilai yang nominal (1 m^3/min 63 detik dan minimal (0.5 m^3/min 75 detik.

  4. The influence of annealing temperature on the strength of TRISO coated particles

    Energy Technology Data Exchange (ETDEWEB)

    Rooyen, I.J. van, E-mail: Isabel.vanrooyen@pbmr.co.z [Pebble Bed Modular Reactor (Pty) Ltd., 1279 Mike Crawford Avenue, Centurion (South Africa); Department of Physics, Nelson Mandela Metropolitan University, Port Elizabeth (South Africa); Neethling, J.H. [Department of Physics, Nelson Mandela Metropolitan University, Port Elizabeth (South Africa); Rooyen, P.M. van [Pebble Bed Modular Reactor (Pty) Ltd., 1279 Mike Crawford Avenue, Centurion (South Africa)

    2010-07-31

    The integrity of the Pebble Bed Modular Reactor (PBMR) fuel, and specifically the SiC layer system of the Tristructural Isotropic (TRISO) coated particle (CP), namely inner pyrolytic carbon, silicon carbide and outer pyrolytic carbon (I-PyC-SiC-O-PyC), determines the containment of fission products. The PBMR fuel consists of TRISO coated particles (CPs) embedded in a graphite matrix. One of the characterization techniques investigated by PBMR is the determination of strength of CPs. It is a well known metallurgical fact that temperature, amongst many other parameters, may influence the strength of a material. A recently developed method for measuring the strength of the TRISO coated particles was used and is briefly described in this article. The advantages of this method are demonstrated by the comparison of strength measurements of five experimental PBMR CP batches as a function of annealing temperature. Significant modification of strength after annealing was measured with increased temperature within the range 1000-2100 {sup o}C. The interesting feature of decreasing standard deviation of the strength with increasing temperature will also be discussed with a possible explanation. A significant difference in coated particle strength is also demonstrated for two CP batches with layer thickness on the extremities of the SiC layer thickness specification. The effect of long duration annealing on these strength values will also be demonstrated by comparing results from 1 h to 100 h annealing periods of coated particles at a temperature of 1600 {sup o}C.

  5. Pengolahan Sampah Secara Pirolisis dengan Variasi Rasio Komposisi Sampah dan Jenis Plastik

    Directory of Open Access Journals (Sweden)

    Qonita Rachmawati

    2015-03-01

    Full Text Available Pada tahun 2012, sampah yang dihasilkan sebesar 1200 ton/hari. Jika permasalahan sampah di Kota Surabaya tidak ditangani dengan baik maka akan menimbulkan beberapa masalah antara lain: masalah kesehatan dan masalah kebersihan. Oleh karena itu, diperlukan metode yang dapat mengolah sampah namun tidak menimbulkan masalah baru lainnya. Salah satu metode pengolahan sampah yang telah dikembangkan, yaitu metode pirolisis. Penelitian ini bertujuan untuk menentukan pengaruh jenis plastik dan komposisi terhadap produk hasil pirolisis. Pada penelitian ini digunakan reaktor dengan kapasitas 500 g yang berbahan stainless steel. Variabel yang digunakan yaitu jenis sampah plastik dan komposisi sampah. Jenis sampah yang digunakan yaitu sampah plastik HDPE (High Density Polyethylene, PET (Poly Ethylene Terephthalate, dan PS (Poly Styrene. Komposisi sampah yang digunakan antara lain: 100:0, 75:25, dan 50:50. Temperatur yang digunakan pada reaktor yaitu 500°C dengan waktu 30 menit. Penelitian dimulai dari persiapan bahan uji, persiapan reaktor, percobaan pendahuluan, dan penelitian dengan reaktor pirolisis. Selanjutnya dilakukan analisis untuk masing-masing hasil produk. Penelitian ini jenis sampah plastik yang menghasilkan gas tertinggi yaitu jenis plastik PET sebesar 45,40% dan jenis plastik yang menghasilkan wax tertinggi yaitu jenis plastik HDPE sebesar 69,91%. Sedangkan komposisi yang menghasilkan gas tertinggi yaitu komposisi dengan ranting 25% dan PET 75% sebesar 71,24% dan komposisi yang menghasilkan wax tertinggi yaitu komposisi dengan ranting 25% dan PS 75% sebesar 61,36%.

  6. ANALISA KARAKTERISTIK MINYAK PLASTIK HASIL DUA KALI PROSES PIROLISIS

    Directory of Open Access Journals (Sweden)

    Untung Surya Dharma

    2015-06-01

    Full Text Available Limbah plastik dapat dimanfaatkan sebagai bahan baku minyak plastik dengan menggunakan proses pirolisis. Minyak plastik yang dihasilkan dapat dimanfaatkan sebagai zat aditif atau campuran bahan bakar pada mesin. Pada Penelitian ini, proses pembuatan minyak plastik menggunakan dua kali proses pirolisis. Suhu reaktor pada proses pirolisis yang pertama dan kedua berbeda berturut-turut yaitu 200 oC dan 150 oC. Dari hasil penelitian ini diketahui bahwa pada proses pirolisis pertama dengan suhu reaktor 200 oC, dari 25 kg bahan baku menghasilkan 15,5 liter minyak plastik dalam waktu 80 jam. Sedangkan pada proses pirolisis kedua dengan suhu reaktor 150 oC, dari 15 liter minyak plastik dari hasil proses pirolisis pertama menghasilkan 11,6 liter minyak plastik dalam waktu 3,33 Jam. Adapun karakter minyak plastik yang dihasilkan adalah massa jenis 771,4 kg/m3, Viskositas 0,501 m2/s dan Nilai kalor 10518 kJ/kg

  7. Effect of basal metabolic rate on the bone mineral density in middle to old age women in Taiwan.

    Science.gov (United States)

    Hsu, Wei-Hsiu; Fan, Chun-Hao; Lin, Zin-Rong; Hsu, Robert Wen-Wei

    2013-09-01

    Basal metabolic rate (BMR) reflects a combination of cardiopulmonary function and lean body mass resulting from regular physical activity. Though many studies have examined the relationships between bone mineral density (BMD) and body composition, little is known regarding the relationship between BMD and BMR. The purpose of this study was to investigate the relationship between BMR, anthropometric parameters, body composition and BMD in postmenopausal women in Taiwan. Two hundred and eighty-nine women between the ages of 40 and 80 years were included in this cross-sectional study. The following parameters were assessed: height, body weight, total body fat (TBF), BMR, waist-to-hip ratio, grip strength, and back strength. Differences in all variables between osteoporotic and non-osteoporotic women (categorized according to decades in age) were calculated using a one-way analysis of variance (ANOVA) and a Bonferroni post-hoc test. Multiple linear regression with a backward stepwise approach was performed to evaluate the relationship between these measurements and BMD. Among women over 50 years of age, those who were non-osteoporotic had higher BMR, BMI, and body fat by comparison to their osteoporotic counterparts (pBMR and body fat significantly predicted BMD of the femoral neck (adjusted beta coefficients of 0.304 and 0.190, respectively; pBMR and body fat also predicted an increased vertebral BMD (adjusted beta coefficients of 0.310 and 0.141, respectively; pBMR is closely associated with BMD in elderly persons, and may be a novel target for interventions aimed at preventing the age-related decline in BMD. Copyright © 2013 Elsevier Ireland Ltd. All rights reserved.

  8. Pra Desain Pabrik Sorbitol dari Tepung Tapioka dengan Hidrogenasi Katalitik

    Directory of Open Access Journals (Sweden)

    Hellen Kartika Dewi

    2014-03-01

    Full Text Available Sorbitol yang dikenal juga sebagai glusitol, adalah suatu gula alkohol yang dimetabolisme lambat di dalam tubuh. Sorbitol banyak digunakan sebagai bahan baku untuk industri barang konsumsi dan makanan seperti pasta gigi, permen, kosmetika, farmasi, vitamin C, termasuk industri tekstil dan kulit. Pembuatan sorbitol dari bahan baku tepung tapioka. Pabrik sorbitol ini direncanakan akan didirikan di Propinsi Jawa Tengah tepatnya di Kabupaten Batang dengan kapasitas produksi 30.000 ton/tahun. Proses produksi Sorbitol menggunakan proses hidrogenasi katalitik. Pembuatan sorbitol dari bahan baku pati melalui dua tahap proses utama yaitu proses perubahan starch menjadi glukosa melalui hidrolisa double enzym. Enzim yang digunakan yaitu α-amylase dan glukoamylase. Proses hidrogenasi katalitik dilakukan dengan mereaksikan larutan dekstrose dan gas hidrogen bertekanan tinggi dengan menambahkan katalis nikel dalam reaktor (Reaktor Hidrogenasi. Gas hidrogen masuk dari bawah reaktor secara bubbling dan larutan dekstrose diumpankan dari atas reaktor sehingga kontak yang terjadi semakin baik. Sorbitol yang di hasilkan dalam pradesain pabrik sorbitol ini dengan konsentrasi 58,2%. Pendirian pabrik sorbitol memerlukan biaya investasi modal tetap (fixed capital sebesar Rp 168.801.192.952, modal kerja (working capital  Rp 29.788.445.815, investasi total Rp 198.589.638.767, Biaya produksi per tahun Rp 368.832.813.809 dan  hasil penjualan per tahun Rp 540.000.078.750. Dari analisa ekonomi didapatkan BEP sebesar 26,32%. ROI sesudah pajak 48,5 %, POT sesudah pajak 2,14 tahun. Dari segi teknis dan ekonomis, pabrik ini layak untuk didirikan.

  9. RA Research reactor, Annual report 1972; Istrazivacki nuklearni reaktor RA - Izvestaj za 1972. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al. [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1972-12-15

    During 1972, the total production was 31151 MWh which is 3.8% higher than planned. The reactor was used for irradiation and experiments according to the demand of 381 users, of which 340 from the Institute and 41 external users. This report contains detailed data about reactor power and experiments performed in 1972. Discrepancies from the action plan, meaning higher production was achieved due to special demands of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to announced power cuts. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. There were no longer interruptions. One shorter interruption (shorter than 24 hours) caused by removal of a UO{sub 2} capsule from the core, placed there for measuring heat transfer. Total personnel exposure dose was lower than during previous years. One accident caused contamination with gases and aerosols containing mainly shot-living isotopes. Decontamination od surfaces was less than during previous years. Practically there was no surface contamination that would demand action of the decontamination team, except for the regular decontamination after refueling. It was concluded that the successful operation in 1972 has a special significance having taking in account the financial crisis caused by the unresolved status of the reactor. It is emphasised, in the plan for the next year that there is an urgent need of making a long-term plan of rector application. It is indispensable to finish preparatory tasks for replacing the fuel with the highly enriched fuel elements by 1974, and building the core emergency cooling system. [Serbo-Croat] Ukupni rad Reaktora RA je u 1972. godini iznosio je 31151 MWh odnosno 3,8% vise od planiranog. Reaktor je koriscen za ozracivanja i eksperimente za 381 korisnika od cega 340 iz Instituta i 41 za korisnike izvan Instituta. Ovaj izvestaj sadrzi detaljne podatke o radu i

  10. Integrating 3D CAD data for manufacturing and fabrication the core model of reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Abu Bakar Harun

    2005-01-01

    This paper describe the intrigue integration of digital 3 Dimensional Computer Aided Design (3D CAD) data manipulation for the Core Model fabrication of REAKTOR TRIGA PUSPATI and ready for mass manufacturing. 3 Dimensional CAD data from Computer Aided Design program will be used as an interpreter in the fabrication of this project. The Core Model of REAKTOR TRIGA PUSPATI will be fabricated with the aid of 3D CAD drawings and digital files. The components will be segregated and divided into 2 categories namely Conventional d Rapid Fabrication. (Author)

  11. Analysis of a 115MW, 3 shaft, helium Brayton cycle

    International Nuclear Information System (INIS)

    Pradeepkumar, K.N.

    2002-01-01

    This research theme is originated from a development project that is going on in South Africa, for the design and construction of a closed cycle gas turbine plant using gas-cooled reactor as the heat source to generate 115 MW of electricity. South African Power utility company, Eskorn, promotes this developmental work through its subsidiary called PBMR (Pebble Bed Modular Reactor). Some of the attractive features of this plant are the inherent and passive safety features, modular geometry, small evacuation area, small infrastructure requirements for the installation and running of the plant, small construction time, quick starting and stopping and also low operational cost. This exercise is looking at the operational aspects of a closed cycle gas turbine, the finding of which will have a direct input towards the successful development and commissioning of the plant. A thorough understanding of the fluid dynamics in this three-shaft system and its transient performance analysis were the two main objectives of this research work. A computer programme called GTSI, developed by a previous Cranfield University research student, has been used in this as a base programme for the performance analysis. Some modifications were done on this programme to improve its control abilities. The areas covered in the performance analysis are Start-up, Shutdown and Load ramping. A detailed literature survey has been conducted to learn from the helium Turbo machinery experiences, though it is very limited. A critical analysis on the design philosophy of the PBMR is also carried out as part of this research work. The performance analysis has shown the advantage, disadvantage and impact of various power modulation methods suggested for the PBMR. It has tracked the effect of the operations of the various valves included in the PBMR design. The start-up using a hot gas injection has been analysed in detail and a successful start region has been mapped. A start-up procedure is also written

  12. RA Reactor; Reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1978-02-15

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation. [Serbo-Croat] Pored osnovnih karakeristika reaktora RA, organizacije rada i finansijskih pokazatelja, razmatra se stanje opreme reaktora nakon 18 godina rada, pitanja dozvole za rad sa 80% obogacenim gorivom, problem skladistenja isluzenog goriva u bazenu zgrade reaktora i potreba za obnavljanjem komponenti opreme, pre svega elektronske.

  13. RA Reactor; Reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-07-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels. [Serbo-Croat] Ovo (prvo) poglavlje sadrzi sledece: Opis reaktora RA; semu organizacije rada i rukovodjenja; prava i duznosti direktora i rukovodioca pogona reaktora, propise o rezimu rada i kretanja u zgradi reaktora, propise o izvodjenju eksperimenata, propise o unosenju uzoraka u eksperimentalne kanale reaktora.

  14. Pebble bed modular reactors versus other generation technologies. Costs and challenges for South Africa

    International Nuclear Information System (INIS)

    Grubert, Emily; Parks, Brian; Schneider, Erich; Sekar, Srinivas

    2011-01-01

    South Africa is Africa's major economy, with plans to double its electricity generation capacity by 2026. South Africa has spent almost two decades developing a nuclear reactor known as a Pebble Bed Modular Reactor (PBMR), which could provide substantial benefits to the electricity grid but was recently mothballed due to high costs. This work estimates the lifecycle financial costs of South African PBMRs, then compares these costs to those of five other generation options: coal, nuclear as pressurized water reactors (PWRs), wind, and solar as photovoltaics (PV) or concentrating solar power (CSP). Each technology is evaluated with low, base case, and high assumptions for capital costs, construction time, and interest rates. Decommissioning costs, project lifetime, capacity factors, and sensitivity to carbon price are also considered. PBMR could be cost competitive with coal under certain low cost conditions, even without a carbon price. However, international lending practices and other factors suggest that a high capital cost, high interest rate nuclear plant is likely to be competing with a low capital cost, low interest rate coal plant in a market where cost recovery is challenging. PBMR could potentially become more competitive if low rate international loans were available to nuclear projects or became unavailable to coal projects. (author)

  15. APLIKASI TEKNIK AAN DI REAKTOR RSG-GAS PADA PENENTUAN UNSUR ESENSIAL DAN TOKSIK DI DALAM IKAN DAN PAKAN IKAN

    Directory of Open Access Journals (Sweden)

    Saeful Yusuf

    2015-03-01

    Full Text Available Pada makalah ini diuraikan tentang aplikasi teknik AAN (Analisis Aktivasi Neutron dalam penentuan konsentrasi unsur-unsur esensial dan cemaran yang terkandung di dalam beberapa spesies ikan dan pakan ikan. Unsur-unsur esensial yang terkandung dalam pakan ikan buatan juga dianalisis untuk mengetahui pengaruhnya terhadap ikan. Penentuan unsur menggunakan teknik AAN dengan metode perbandingan dan metode k0-AAN. Sampel diiradiasi di reaktor RSG-GAS yang memiliki fluks neutron thermal 5 x 1013 n.cm-2.s-1 pada daya 15 MW. Hasil penelitian menunjukkan bahwa 12 unsur di dalam 11 spesies ikan air laut dan air tawar telah ditentukan yaitu As, Br, Cr, Co, Cs, Fe, Hg, K, Na, Rb, Se and Zn. Konsentrasi cemaran As didalam ikan laut sudah melampaui batas maksimum 1 mg/kg, sedangkan konsentrasi cemaran Hg masih dibawah batas maksimum 0,5 mg/kg, baik untuk ikan laut maupun ikan air tawar. Unsur K dan Na merupakan unsur makroesensial sedangkan unsur Cr, Co, Fe, Se and Zn adalah termasuk unsur mikroesensial. Secara umum ditunjukkan bahwa kandungan mineral didalam ikan laut lebih tinggi konsentrasinya dibandingkan ikan air tawar. Br, Cs dan Rb merupakan unsur-unsur non esensial yang teridentifikasi dalam semua ikan yang dianalisis. Penelitian terhadap pakan ikan air tawar menunjukkan bahwa semua unsur yang teridentifikasi juga terdapat di dalam ikan laut dan ikan air tawar. Hal ini menunjukkan bahwa pakan ikan berkontribusi terhadap konsentrasi unsur di dalam ikan air tawar. Kata kunci : Analisis aktivasi neutron, unsur esensial, unsur cemaran, ikan, pakan ikan   This paper reported on the application of NAA (Neutron Activation Analysis Technique in the determination of the concentration of the essential and toxic elements in some species of fish and fish feed. Determination of elements using instrumental NAA technique with comparison and k0-INAA methods. Samples were irradiated in the RSG-GAS which has a thermal neutron flux  5.0E +13 ncm-2s-1. The results

  16. ANALISIS SKENARIO KEGAGALAN SISTEM UNTUK MENENTUKAN PROBABILITAS KECELAKAAN PARAH AP1000

    Directory of Open Access Journals (Sweden)

    D.T. Sony Tjahyani

    2014-03-01

    Full Text Available Kejadian Fukushima telah menunjukkan bahwa kecelakaan parah dapat terjadi, maka dari itu sangatlah penting untuk menganalisis tingkat keselamatan pada reaktor daya. Berdasarkan rekomendasi expert mission IAEA setelah kejadian Fukushima, perlu dilakukan upaya untuk meminimalisasi terjadinya kecelakaan parah yaitu dengan melakukan proses pendinginan yang maksimal. Dalam konsep keselamatan fasilitas nuklir, khususnya reaktor daya telah diterapkan konsep keselamatan berlapis (Defence in Depth, DiD. Konsep keselamatan tersebut terdiri atas 5 level pertahanan yang bertujuan mencegah dan mengurangi lepasan produk fisi ke masyarakat dan lingkungan pada saat reaktor daya mengalami kecelakaan. Dalam reaktor telah didesain sistem atau tindakan yang mempunyai fungsi untuk mengatasi setiap level tersebut. Tujuan dari analisis ini adalah menentukan probabilitas kecelakaan parah dengan melakukan skenario kegagalan sistem dalam proses pendinginan di reaktor. Sebagai obyek analisis adalah reaktor daya AP1000, karena jenis reaktor ini sedang banyak dibangun saat ini. Skenario dilakukan dengan mengasumsikan beberapa kombinasi kegagalan sistem yang termasuk dalam DiD level 2 dan 3. Kegagalan sistem kemudian dianalisis dengan menggunakan analisis pohon kegagalan berdasarkan perangkat lunak SAPHIRE ver. 6.76. Dari analisis didapatkan probabilitas gagal dari kelompok sistem DiD level 2 dan 3 pada AP1000 masih di bawah batas kriteria dari IAEA yaitu lebih kecil dari 10-2, serta probabilitas kecelakaan parah didapatkan sebesar 6,17 x 10-10. Berdasarkan analisis ini disimpulkan bahwa AP1000 mempunyai tingkat keselamatan yang cukup tinggi, karena melalui skenario kegagalan sistem didapatkan probabilitas kecelakaan parah yang sangat kecil.   ABSTRACT Fukushima accident has shown that severe accident could be occurred, therefore it is important to analyze safety level of nuclear power plants. Based on the recommendations of IAEA expert mission after the Fukushima accident

  17. A CFD Study on Inlet Plenum Flow Field of Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Lee, Won Jae; Chang, Jong Hwa

    2005-01-01

    High temperature gas cooled reactor, largely divided into two types of PBR (Pebble Bed Reactor) and PMR (Prismatic Modular Reactor), has becomes great interest of researchers in connection with the hydrogen production. KAERI has started a project to develop the gas cooled reactor for the hydrogen production and has been doing in-depth study for selecting the reactor type between PBR and PMR. As a part of the study, PBMR (Pebble Bed Modular Reactor) was selected as a reference PBR reactor for the CFD analysis and the flow field of its inlet plenum was simulated with computational fluid dynamics program CFX5. Due to asymmetrical arrangement of pipes to the inlet plenum, non-uniform flow distribution has been expected to occur, giving rise to non-uniform power distribution at the core. Flow fields of different arrangement of inlet pipes were also investigated, as one of measures to reduce the non-uniformity

  18. Activity of the RA Reactor Physics group in 1980 - Definition of the Operation conditions for future safe and economical RA reactor operation with 80% enriched fuel; Prilog Ia - Rad sluzbe za fiziku reaktora RA u 1980. godini - Definisanje pogonskih uslova za dalji siguran i ekonomican rad reaktora RA sa 80% obogacenim gorivom

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1980-12-15

    During 1980. the RA reactor was not in operation. That is why this period was devoted to definition of operating conditions for further reactor operation with 80% enriched fuel. The fuel elements which were in the core at the moment of shutdown in March 1979will not be used again (388 80% enriched fuel elements, and 511 2% enriched fuel elements). The reactor will be operated only with 80% enriched fuel, staring with initiat core configuration with 440 elements on the borders gradually changing to equi;librium core with 720 fuel elements. The analyses were concerned with safety issues of future operation. [Serbo-Croat] Reaktor RA tokom 1980 godine nije radio. Zbog toga je ovaj period iskoriscen za intenzivan rad na definisanju pogonskih uslova za dalji rad reaktora sa 80% obogacenim gorivom. Gorivo sa kojim je reaktor prestao da radi marta 1979. godine (388 gorivnih elemenata 80% obogacenoh i 511 elementa 2% obogaceno) nece biti vraceno u jezgro, vec ce reaktor raditi samo sa 80% obogacenim gorivom, pocev od konfiguracije sa 440 elemenata na periferiji do ravnotezne konfiguracije sa 720 elemenata. Sve nalize su se bavile sigurnosnim aspektima rada reaktora sa visokoobogacenim gorivom u jezgru.

  19. Kajian Pemilihan Sumber Mikroorganisme Solid Phase Microbial Fuel Cell (SMFC Berdasarkan Jenis dan Volume Sampah, Power Density dan Efisiensi Penurunan COD

    Directory of Open Access Journals (Sweden)

    Ganjar Samudro

    2017-06-01

    Full Text Available Mikroorganisme merupakan salah satu komponen penting dalam proses Solid Phase Microbial Fuel Cell (SMFC untuk degradasi bahan organik dan transfer elektron. Pemilihan sumber mikroorganisme menjadi metode yang paling sederhana untuk dikaji sebagai informasi awal ketersediaan dan identifikasi jenis mikroorganisme yang mendukung proses SMFC. Tujuan kajian ini adalah untuk memilih sumber mikroorganisme tanah, septic tank dan sedimen sungai yang tepat digunakan dalam proses SMFC berdasarkan jenis dan volume sampah, power density, dan efisiensi penurunan COD. Kajian ini didasarkan pada hasil penelitian menggunakan reaktor SMFC tipe single chamber microbial fuel cell dengan variabel jenis dan volume sampah , serta sumber mikroorganisme. Metode perbandingan secara kuantitatif dilakukan berdasarkan kecenderungan nilai power density dan efisiensi penurunan COD tertinggi di antara jenis dan volume sampah kantin, dedaunan dan komposit kantin-dedaunan. Hasil yang didapatkan adalah sumber mikroorganisme tanah dan sedimen sungai tepat digunakan untuk volume sampah 1/3 dan 2/3 dari volume reaktor, sedangkan sumber mikroorganisme septic tank tepat digunakan untuk volume sampah 1/3 dan 1/2 dari volume reaktor. Sumber mikroorganisme dari septic tank menunjukkan kinerja power density dan efisiensi penurunan COD yang lebih rendah dibandingkan sumber mikroorganisme tanah dan sedimen sungai.

  20. Penurunan Logam Timbal (Pb pada Limbah Cair TPA Piyungan Yogyakarta dengan Constructed Wetlands Menggunakan Tumbuhan Eceng Gondok (Eichornia Crassipes

    Directory of Open Access Journals (Sweden)

    Eko Siswoyo

    2015-10-01

    Full Text Available Salah satu permasalahan lingkungan yang ditimbulkan dari adanya lindi di TPA Piyungan yaitu pencemaran pada badan air, sungai dan air tanah. Untuk mengatasi permasalahan ini salah satunya dengan sistem Constructed Wetlands dengan menggunakan tumbuhah eceng gondok. Tujuan dari penelitian ini adalah untuk mengetahui tingkat penurunan konsentrasi Timbal (Pb yang terdapat dalam limbah cair TPA Piyungan dengan Constructed Wetlands menggunakan tumbuhan eceng gondok dan untuk mengetahui seberapa besar kapasitas serapan tumbuhan eceng gondok terhadap kandungan Timbal (Pb dalam limbah cair TPA Piyungan.Dalam penelitian ini digunakan reaktor yang terbuat dari kayu yang dilapisi plastik dengan ukuran 0,5 m x 1,0 m. Setiap reaktor diberi media tanah 5 cm, dan diberi tumbuhan sebanyak 14 buah. Reaktor tersebut diberi perlakuan dengan konsentrasi limbah yang bervariasi (100%, 75%, 50%, 25%, dan 0%, dan waktu pengambilan sampel (0, 3, 6, 9, 12 hari. Dengan menggunakan metode SSA (Spektrofotometri Serapan Atom.Berdasarkan pengujian diperoleh bahwa penurunan logam Pb pada limbah cair TPA Piyungan hari ke- 12, yaitu sebesar 0.0501mg/L pada konsentrasi 100%, 0.0295mg/L pada konsentrasi 75%, 0.0267mg/L pada konsentrasi 50% dan 0.0041 mg/L pada konsentrasi 25%.

  1. Nuclear reactor PBMR and cogeneration

    International Nuclear Information System (INIS)

    Ramirez S, J. R.; Alonso V, G.

    2013-10-01

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  2. The analytic nodal method in cylindrical geometry

    International Nuclear Information System (INIS)

    Prinsloo, Rian H.; Tomasevic, Djordje I.

    2008-01-01

    Nodal diffusion methods have been used extensively in nuclear reactor calculations, specifically for their performance advantage, but also for their superior accuracy. More specifically, the Analytic Nodal Method (ANM), utilising the transverse integration principle, has been applied to numerous reactor problems with much success. In this work, a nodal diffusion method is developed for cylindrical geometry. Application of this method to three-dimensional (3D) cylindrical geometry has never been satisfactorily addressed and we propose a solution which entails the use of conformal mapping. A set of 1D-equations with an adjusted, geometrically dependent, inhomogeneous source, is obtained. This work describes the development of the method and associated test code, as well as its application to realistic reactor problems. Numerical results are given for the PBMR-400 MW benchmark problem, as well as for a 'cylindrisized' version of the well-known 3D LWR IAEA benchmark. Results highlight the improved accuracy and performance over finite-difference core solutions and investigate the applicability of nodal methods to 3D PBMR type problems. Results indicate that cylindrical nodal methods definitely have a place within PBMR applications, yielding performance advantage factors of 10 and 20 for 2D and 3D calculations, respectively, and advantage factors of the order of 1000 in the case of the LWR problem

  3. State space model extraction of thermohydraulic systems – Part II: A linear graph approach applied to a Brayton cycle-based power conversion unit

    International Nuclear Information System (INIS)

    Uren, Kenneth Richard; Schoor, George van

    2013-01-01

    This second paper in a two part series presents the application of a developed state space model extraction methodology applied to a Brayton cycle-based PCU (power conversion unit) of a PBMR (pebble bed modular reactor). The goal is to investigate if the state space extraction methodology can cope with larger and more complex thermohydraulic systems. In Part I the state space model extraction methodology for the purpose of control was described in detail and a state space representation was extracted for a U-tube system to illustrate the concept. In this paper a 25th order nonlinear state space representation in terms of the different energy domains is extracted. This state space representation is solved and the responses of a number of important states are compared with results obtained from a PBMR PCU Flownex ® model. Flownex ® is a validated thermo fluid simulation software package. The results show that the state space model closely resembles the dynamics of the PBMR PCU. This kind of model may be used for nonlinear MIMO (multi-input, multi-output) type of control strategies. However, there is still a need for linear state space models since many control system design and analysis techniques require a linear state space model. This issue is also addressed in this paper by showing how a linear state space model can be derived from the extracted nonlinear state space model. The linearised state space model is also validated by comparing the state space model to an existing linear Simulink ® model of the PBMR PCU system. - Highlights: • State space model extraction of a pebble bed modular reactor PCU (power conversion unit). • A 25th order nonlinear time varying state space model is obtained. • Linearisation of a nonlinear state space model for use in power output control. • Non-minimum phase characteristic that is challenging in terms of control. • Models derived are useful for MIMO control strategies

  4. Increased BMR in overweight and obese patients with type 2 diabetes may result from an increased fat-free mass.

    Science.gov (United States)

    Sun, Min-xian; Zhao, Shi; Mao, Hong; Wang, Zhong-jing; Zhang, Xu-yan; Yi, Lan

    2016-02-01

    The study aimed to determine the relationships between the basal metabolic rate (BMR) and body composition of overweight and obese Chinese adults with type 2 diabetes mellitus (DM). This cross-sectional clinical study enrolled 193 Chinese adults with type 2 DM who were overweight (24 kg/m(2)=BMI≤28 kg/m(2), n=99), or obese (BMI ≥28 kg/m(2), n=94). Ninety-seven adults with normal BMIs, including 50 DM patients and 47 healthy adults, were recruited as a control group. BMR was measured by indirect calorimetry; predicted BMR was calculated according to the Schofield equation; and the relationships between BMR, body composition, and biochemical results were determined by the Pearson correlation. The results showed that obese DM patients had significantly higher BMRs than both overweight patients (PBMR was significantly lower than the predicted BMR (PBMR in both DM patients (r=0.874, P<0.01) and in healthy controls (r=0.902, P<0.01). It was concluded that overweight and obese Chinese adults with type 2 DM had increased BMRs compared with normal-weight controls, which may result from the difference in fat-free mass.

  5. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    International Nuclear Information System (INIS)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-01-01

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning

  6. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Task 5 Report: Generation IV Reactor Virtual Mockup Proof-of-Principle Study

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Task 5 report is part of a 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Created a virtual mockup of PBMR reactor cavity and discussed applications of virtual mockup technology to improve Gen IV design review, construction planning, and maintenance planning.

  7. Wpływ czasu napowietrzania na pracę reaktora SBR i SBBR

    Directory of Open Access Journals (Sweden)

    Małgorzata Makowska

    2016-06-01

    Full Text Available W pracy przedstawiono efekty oczyszczania małych ilości ścieków w bioreaktorach porcjowych. Przebadano dwa równolegle pracujące reaktory, z czego jeden był klasycznym systemem SBR, a drugi to reaktor SBBR ze złożem ruchomym. Obydwa ciągi technologiczne oczyszczały taką samą ilość ścieków bytowych wstępnie podczyszczonych w piaskowniku. Trzy kolejne serie badań różniły się długością napowietrzania. Porównano pracę reaktorów w kolejnych seriach oraz analizowano wyniki uzyskane w obu systemach. Największą skuteczność usuwania związków węgla organicznego (jako ChZT i azotu (jako N-NH4 uzyskano w reaktorze SBBR w trzeciej serii badań (średnio odpowiednio 90 i 62%. Reaktor SBBR pracował bardziej stabilnie oraz usuwał na drodze biologicznej fosfor z największą średnią skutecznością 83% w serii trzeciej. Ścieki oczyszczone w tym reaktorze charakteryzowały się mniejszym stężeniem zawiesiny, co świadczy o skuteczniejszej pracy reaktora z wypełnieniem w fazie sedymentacji.

  8. Põhja-Korea lubas taas jätta tuumaplaanid / Jürgen Tamme

    Index Scriptorium Estoniae

    Tamme, Jürgen

    2007-01-01

    Pekingis toimunud kuuepoolsete desarmeerimiskõneluste tulemusena on Põhja-Korea nõus andma ülevaate riigi tuumarajatistest ning alustama nende likvideerimist. 14. juulil suleti Yongbyoni tuumakeskuse reaktor. Lisa: Tuumavaidlus

  9. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987; Istrazivacki nuklearni reaktor RA Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1987. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1987-12-15

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction. [Serbo-Croat] Reaktor RA nije radio usled zabrane Izvsnog veca Skupstine Srbije od 27. avgusta 1984. U cilju povecanja pouzdanosti rada reaktora a da bi se udovoljilo zakonskim propisima sto je uslov za dobijanje stalne dozvole za rad realizovana su tri velika zahvata na reaktoru RA. Ovi zahvati obuhvatili su izgradnju sistema za hladjenje jezgra reaktora u slucaju nuzde, rekonstukciju postojeceg sistema specijalne ventilacije i rekonstrukciju sistema napajanja elektricnom energijom neophodnih potrosaca reaktora RA. Zapoceti su radovi na modernizaciji intrumentacije reaktora RA, projekat je izradjen u sovjetskoj organizaciji Atomenergoeksport, a trebalo bi da se realizuje do kraja 1989. godine. U cilju povecanja prostora za skladistenje ozracenog nuklearnog goriva i njegovog efikasnijeg koriscenja, izradjen je su projekti za rekonstrukciju postojecih uredjaja za rukovanje gorivom, povecanje smestajnog kapaciteta i preciscavanje vode u bazenima za odlezavanje. Realizaija ovih

  10. Adamkus : EU might be willing to negotiate on postponing Ignalina reactor closure

    Index Scriptorium Estoniae

    2008-01-01

    President Valdas Adamkus teatas võimalusest, et EL alustab arutelu Ignalina tuumaelektrijaama sulgemise edasilükkamisest. Uue jaama valmimise aeg pole teada ning praegune reaktor on tänu rootslaste abile turvaline

  11. Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

    International Nuclear Information System (INIS)

    In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

    2008-01-01

    Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

  12. Risk-informed design of a pebble bed gas reactor

    International Nuclear Information System (INIS)

    Ritterbusch, Stanley; Dimitrijevic, Vesna; Simic Zdenko; Savkina Marina

    2003-01-01

    One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor design technology and operating experience. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past experience, the pre-application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional 'risk-informed' improvements to the licensing process. These improvements are aimed at resolving new design and regulatory issues using a plant-wide integrated evaluation method - state-of-the-art Probabilistic Risk Assessment - which addresses all significant design features and operating modes. The integrated PRA evaluation is supported by the usual deterministic design analyses, engineering judgments, and margins added to address uncertainties (i.e., defense-in-depth). The work performed for this paper was completed as part of the United States Department of Energy's Nuclear Energy Research Initiative. The purpose of this particular project was to develop the methods for a new 'highly risk-informed' design and regulatory process. In this work. PRA techniques were applied in order to provide an integrated and systematic analysis of the plant design, to quantify uncertainties and explicitly account for defense-in-depth features. This work concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The implementation example completed for this project included specification of the design configuration, use of the PRA to evaluate the design, and iterations to identify design changes that improve the overall level of safety and system reliability. This paper summarizes the new 'highly risk-informed' design process, the design of the PBMR, and the results obtained. These results, consistent with the known inherent safety features of a pebble

  13. Power generation in Southern Africa

    International Nuclear Information System (INIS)

    Beer, J.A. de

    2002-01-01

    This paper outlines the main characteristics of power generation in Southern Africa, in terms of primary energy resources, existing and projected power supply and demand, types and location of power plants, regional integration, and environmental management aspects. Various options for future development of power generation are presented as part of an overall integrated resource planning (IRP) process for the power industry. These include coal and natural gas based options, hydro power and other renewable energy, and nuclear power plants. A specific option, the pebble bed modular reactor (PBMR), under development by Eskom Enterprises and other international and local partners, is described in terms of overall design parameters, inherent safety features, economics and environmental aspects. Included is a high level discussion on the selection of materials for the design of this PBMR plant, an advanced design version of a high temperature gas reactor (HTGR). (orig.)

  14. Preliminary studi on neutronic aspect of a conceptual design of the Kartini reactor base ADS facility

    International Nuclear Information System (INIS)

    Tegas Sutondo

    2012-01-01

    A preliminary study on neutronic aspect of a conceptual design of ADS facility with the basis of Kartini Reaktor, has been performed. The study was intended to see the feasibility from neutronic point of view of Kartini reactor, to be used as a small scale of NPP’s waste transmutation experimental facility. A SRAC code was used as the basis of calculations. The results indicate that the presence of minor actinides (MA) will give a positive reactivity, which tends to increase with the increase of MA concentrations. Based on the defined criteria of subcriticality and by considering the core power distributions and the level of reactivity contribution of MA element, it is concluded that Kartini reactor is potential enough to be used as an ADS experimental facility, mainly for MA concentration between 30 to 50 % of the assumed mixture of C-MA matrix. (author)

  15. Pushbroom microwave radiometer results from HAPEX-MOBILHY

    International Nuclear Information System (INIS)

    Nichols, W.E.; Cuenca, R.H.; Schmugge, T.J.; Wang, J.R.

    1993-01-01

    The NASA C-130 remote sensing aircraft was in Toulouse, France from 25 May through 4 July 1986, for participation in the HAPEX-MOBILHY program. Spectral and radiometric data were collected by C-130 borne sensors in the visible, infrared, and microwave wavelengths. These data provided information on the spatial and temporal variations of surface parameters such as vegetation indices, surface temperature, and surface soil moisture. The Pushbroom Microwave Radiometer (PBMR) was used to collect passive microwave brightness temperature data. This four-beam sensor operates at the 21-cm wavelength, providing cross-track coverage approximately 1.2 times the aircraft altitude. Observed brightness temperatures for the period were high, ranging from above 240 K about 290 K. Brightness temperature images appeared to correspond well to spatial and temporal soil moisture variation. Previous research has demonstrated that an approximately linear relationship exists between the surface emissivity and surface soil moisture. For these data, however, regression analysis did not indicate a strong linear relationship (r 2 = 0.32 and r 2 = 0.42 respectively) because of the limited range of soil moisture conditions encountered and the small number of ground measurements. When results from wetter soil conditions encountered in another experiment were included, the regression improved dramatically. Based on similar research with the PBMR and an understanding of the ground data collection program, this result was examined to produce recommendations for improvements to future passive microwave research and data collection programs. Examples of surface soil moisture maps generated with PBMR data are presented which appear to be representative of the actual soil moisture conditions

  16. Characteristic behaviour of Pebble Bed High Temperature Gas-cooled Reactors during water ingress events

    International Nuclear Information System (INIS)

    Khoza, Samukelisiwe N.; Serfontein, Dawid E.; Reitsma, Frederik

    2014-01-01

    in order to curb the increase in the multiplication factor, the level of under-moderation in the fuel can be decreased by reducing the heavy metal loading of the fuel spheres and/or the porosity of the graphite reflector block must be increased in order to increase water ingress into the reflectors. For the PBMR-400 the diameter of the central reflector can also be reduced. Results from past studies of water ingress into Pebble Bed Reactors were used to validate and verify the present simulation approach and results. (author)

  17. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988; Istrazivacki nuklearni reaktor RA, deo 1, pogon i odrzavanje nukleanog reaktora RA u 1988. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Badrljica, R; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1988-12-15

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues. [Serbo-Croat] Prema planu za 1988. godinu, reaktor RA je trebalo da pusten u rad oktobra meseca, medjutim nije dobio dozvolu za nastavak rada. Kontrola i odrzavanje opreme izvrsavani su redovno i efikasno, u granicama koje su diktirane raspolozivoscu repromaterijala i rezervnih delova. Najvecu poteskocu pricinjavalo je odrzavanje instrumentacije. Period stajanja u 1988. godini iskoriscen je za remont teskovodnih pumpi u primarnom kolu hladjenja. U cilju povecanja pouzdanosti rada reaktora zapoceti su radovi na modernizaciji instrumentacije, projekat je izradjen u sovjetskoj organizaciji Atomenergoeksport, sklopljen je ugovor o izradi ove opreme koja bi trebalo da bude isporucena do kraja 1990. U cilju povecanja prostora za skladistenje ozracenog

  18. FACTORS INFLUENCING HUMAN RELIABILITY OF HIGH TEMPERATURE GAS COOLED REACTOR OPERATION

    Directory of Open Access Journals (Sweden)

    Sigit Santoso

    2016-10-01

    ABSTRAK Peran dan tindakan operator pada reaktor berpendingin gas akan berbeda dengan peran operator pada operasi tipe reaktor lain. Analisis unjuk kerja operator dan faktor yang berpengaruh dapat dilakukan secara komprehensif melalui analisis keandalan manusia(HRA. Melalui HRA dampak dari kesalahan manusia pada sistem maupun cara untuk mengurangi dampak dan frekuensi kesalahan dapat diketahui. Makalah membahas faktor yang berpengaruh pada tindakan operator, yaitu pada kejadian kecelakaan pendingin reaktor gas bersuhu tinggi-HTGR. Analisis untuk kualifikasi faktor pembentuk kinerja(PSF dilakukan berdasarkan kurva keandalan fungsi waktu, dan metode keandalan manusia yang dikembangkan berdasar pada aspek kognitif yaitu Cognitive Reliability and Error Analysis Method (CREAM. Hasil analisis berdasar kurva keandalan fungsi waktu menunjukkan komponen waktu berkontribusi positif pada peningkatan keandalan operator (PSF<1 pada kondisi semua fitur keselamatan berfungsi sesuai rancangan. Sedangkan pada metoda analisis dengan pendekatan kognitif CREAM diketahui selain faktor ketersediaan waktu, faktor pelatihan dan rancangan HMI juga berkontribusi meningkatkan keandalan operator. Faktor pembentuk kinerja keseluruhan diketahui sebesar 0,25 dengan faktor kontribusi positif dominan atau berpengaruh pada penurunan kesalahan manusia adalah ketersediaan waktu (PSF=0,01, dan faktor kontribusi negatif dominan adalah prosedur dan siklus kerja (PSF=5. Nilai PSF tersebut sebagai faktor pengali dalam perhitungan probabilitas kesalahan manusia. Analisis faktor pembentuk kinerja perlu dikembangkan pada skenario kejadian lain untuk selanjutnya digunakan untuk perhitungan dan analisis keandalan manusia yang komprehensif dan perancangan sistem interaksi manusia mesin di ruang kendali. Kata kunci: PSF, HTGR, operator, ruang kendali, keandalan manusia

  19. Development of a system code for transient analysis in a HTGR

    International Nuclear Information System (INIS)

    Lee, Tae Beom

    2004-02-01

    A GAMMA (GAs Multi-component Multi-dimensional Analysis) code is developed for transient analysis and air ingress analysis in High Temperature Gas-cooled Reactors (HTGR). The PBMR of ESKOM is selected as a reference plant for the High Temperature Gas-cooled Reactor here, which uses a direct helium cycle and pebble fuel. Physical models included in GAMMA are the pebble conduction model, radiation heat transfer model, point kinetics model, decay heat model, and component models for break flow, valve, pump, cooler, power conversion unit model. The temperature distribution and the flow distribution of the PBMR are calculated for initial and accident core in the present study. In the accident analysis, typical design basis accident (DBA), including the load transient accident and depressurization accident into the system are selected and analyzed in detail. The predictions by GAMMA for PBMR at 100% power are compared with those by VSOP and PBR S IM. It turns out that the temperature in the upper region in the third channel predicted by GAMMA is about 62 .deg. C at maximum higher than that by VSOP, but is pretty close to that by PBR S IM. The center temperature of the fuel shows that that predicted by considering swelling effect is higher than that without swelling effect by about 10 .deg. C. The net efficiency of direct system is higher than that of indirect system due to an effect of the circulator power. The transient capability of GAMMA is validated through analytical solution and PBR S IM analyzing the depressurization (Loss Of Coolant Accident, LOCA) and load transient accident. After the LOCA the system pressure decreases dramatically from 8MPa to 0.4MPa within 2 sec. After the PI (Proportional-plus-Integral) controller senses that the power shaft is over the set-point of 3,600 rpm, the bypass valve makes shaft speed back to the set-point

  20. PENGOLAHAN LIMBAH CAIR INDUSTRI FARMASI FORMULASI DENGAN METODE ANAEROB-AEROB DAN ANAEROB-KOAGULASI

    OpenAIRE

    Farida Crisnaningtyas; Hanny Vistanty

    2016-01-01

    Studi ini membahas mengenai pengolahan limbah cair industri farmasi dalam skala laboratorium dengan menggunakan konsep anaerob-kimia-fisika dan anaerob-aerob. Proses anaerob dilakukan dengan menggunakan reaktor Upflow Anaerobic Sludge Bed reactor (UASBr) pada kisaran OLR (Organic Loading Rate) 0,5 – 2 kg COD/m3hari, yang didahului dengan proses aklimatisasi menggunakan substrat gula. Proses anaerob mampu memberikan efisiensi penurunan COD hingga 74%. Keluaran dari proses anaerob diolah lebih ...

  1. PBMR phase 1 study: Seismic and structural design consideration - An overview of principles

    International Nuclear Information System (INIS)

    Wium, D.J.W.

    1997-01-01

    This paper briefly reviews the principles involved in the planning and design of the proposed facility to cater for seismic and structural loads. The conceptual layout is discussed, as well as the different load characteristics and scenarios. An outline is given of model used to estimate the seismic loads, whereafter the different analytical models are discussed. (author)

  2. Eesti Energia valib Ameerikas tuumajaamale reaktoreid / Andres Reimer

    Index Scriptorium Estoniae

    Reimer, Andres

    2009-01-01

    Eesti Energia liitus rahvusvahelise tuumareaktoriprojektiga IRIS, projekti eesmärk on luua lihtsa konstruktsiooniga, ohutu ja suhteliselt odav reaktor. Jürgen Ligi sõnul saab Eesti oma IRIS-tüüpi tuumajaama juba 2019. aastal. Joonis: Potentsiaalne tuumajaam

  3. Office of Nuclear Regulatory Research summary of advanced reactors activities, June 4, 2001

    International Nuclear Information System (INIS)

    2001-01-01

    Pre-application interactions with potential licensee applicants will help NRC prepare for future submittals, through the development of the infrastructure necessary for licensing application reviews. RES has the lead for non-LWR advanced reactor pre-application initiatives and longer-range new technology initiatives. An advanced reactor group has been formed in REAHFB, and is currently performing a pre-application review of Exelon's Pebble Bed Modular Reactor. Recent industry requests for future pre application interaction include General Atomics' Gas Turbine-Modular Helium Reactor (GT-MHR) and Westinghouse International Reactor Innovative and Secure (IRIS) design. RES advanced reactors activities also include participation as an observer in DOE's Generation IV initiative. Pre-Application review objectives include the development of regulatory guidance, licensing approach, and technology-basis expectations for licensing advanced designs, including identifying significant technology, design, safety, licensing and policy issues that would need to be addressed in the licensing process. The presentation described the pre-application process for the Exelon PBMR. NRC first identifies additional information following topical meetings with Exelon, and Exelon formally documents and submits required topical Information. The staff then develops a preliminary assessment and drafts a response which is followed by stakeholder input and comments at a public workshop. Preliminary assessments are discussed with ACRS and ACNW, and Commission papers are written which provide staff positions and recommendations on proposed policy decisions. Some of the significant areas for the PBMR include: Process Issues, Legal and Financial Issues; Regulatory Framework; Fuel Performance and Qualification; Traditional Engineering Design (e.g, Nuclear, Thermal-Fluid, Materials); Fuel Cycle Safety Areas; PRA, SSC Safety Classification; PBMR Prototype Testing

  4. Identification of the key parameters defining the life of graphite core components

    International Nuclear Information System (INIS)

    Mitchell, M.N.

    2005-01-01

    The Core Structures of a Pebble Bed rector core comprise graphite reflectors constructed from blocks. These blocks are subject to high flux and temperatures as well as significant gradients in flux and temperature. This loading combined with the behaviour of graphite under irradiation gives rise to complex stress states within the reflector blocks. At some point, the stress state will reach a critical level and cracks will initiate within the blocks. The point of crack initiation is a useful point to define as the end of the part's life. The life of these graphite reflector parts in a pebble bed reactor (PBR) core determines the service life of the Core Structures. The replacement of the Core Structures' components will be a costly and time consuming. It is important that the components of the Core Structures be designed for the best life possible. As part of the conceptual design of the Pebble Bed Modular Reactor (PBMR), the assessment of the life of these components was examined. To facilitate the understanding of the parameters that influence the design life of the PBMR, a study has been completed into the effect of various design parameters on the design life of a typical side reflector block. Parameters investigated include: block geometry, material property variations, and load variations. The results of this study are to be presented. (author)

  5. Na velikosti nezáleží - je to fraktální

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2014-01-01

    Roč. 7, červen (2014) Institutional support: RVO:61389021 Keywords : fusion * tokamak * fractal * turbulence * Bohm diffusion Subject RIV: BL - Plasma and Gas Discharge Physics http://3pol.cz/1614-na-velikosti-nezalezi-je-to-fraktalni-aneb-slunecni-termojaderny-reaktor

  6. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Bjornard, Trond; Hockert, John

    2011-01-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC and A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC and A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC and A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR (Pty) and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC and A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR and D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present

  7. South Africa's Pebble Bed Modular Reactor, a new design for our nuclear future

    International Nuclear Information System (INIS)

    Nicholls, David

    2001-01-01

    Power Utilities will in the future need to look at various means of generating power during the 21st century. The demands regarding new generation are challenged by such issues as; costs, time to construct, the add on safety requirements of present day nuclear power plant designs and the emissions generated by fossil fuels as reflected in the Kyoto Protocol. These challenges are also aligned with the deforestation, land decimation and releases of methane gas caused by the so-called 'clean' Hydro power plants in many parts of the world. Presently South Africa is looking at various generation mixes for the future. Although the demand in South Africa is currently lower than the capacity, it is anticipated that new capacity will have to be commissioned by about 2008. Even the moderate growth of 2,5% (as was experienced in our last fiscal year) will result in peak electricity demand exceeding capacity between 2005 and 2010. In addition, Eskom's older power stations reach the end of their design life after 2025. South Africa will, therefore, need to access and use all natural resources to produce the additional 20,000MW of electricity that will be needed by 2025 this will of course include a nuclear option. Throughout the world, it is noted that, along with the environmental issues affecting power generation the real leading issue is cost. South Africa has one of the lowest power costs in the world, based on its abundant low-cost coal. As with other Eskom low cost options such as, coal fired generation situated at the pit-head and imported hydro, the PBMR costs will have to meet these demanding cost targets set by Eskom's existing power plants. However, PBMR is virtually independent of location and the intention is that PBMR costs will be in the order of US 2,0c/kWh. The costs of decommissioning, long-term storage of radioactive waste and insurance are included in these estimates. This cost per unit of electricity produced would, however, be much lower than a coal

  8. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Final Report

    International Nuclear Information System (INIS)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-01-01

    Final report of 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Mockups applied to design review of AP600/1000, Construction planning for AP 600, and AP 1000 maintenance evaluation. Proof of concept study also performed for GenIV PBMR models

  9. Generation IV Nuclear Energy Systems Construction Cost Reductions through the Use of Virtual Environments - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Shaw; Anthony Baratta; Vaughn Whisker

    2005-02-28

    Final report of 3 year DOE NERI-sponsored effort evaluating immersive virtual reality (CAVE) technology for design review, construction planning, and maintenance planning and training for next generation nuclear power plants. Program covers development of full-scale virtual mockups generated from 3D CAD data presented in a CAVE visualization facility. Mockups applied to design review of AP600/1000, Construction planning for AP 600, and AP 1000 maintenance evaluation. Proof of concept study also performed for GenIV PBMR models.

  10. HYBRID SULFUR PROCESS REFERENCE DESIGN AND COST ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.; Summers, W.; Boltrunis, C.; Lahoda, E.; Allen, D.; Greyvenstein, R.

    2009-05-12

    This report documents a detailed study to determine the expected efficiency and product costs for producing hydrogen via water-splitting using energy from an advanced nuclear reactor. It was determined that the overall efficiency from nuclear heat to hydrogen is high, and the cost of hydrogen is competitive under a high energy cost scenario. It would require over 40% more nuclear energy to generate an equivalent amount of hydrogen using conventional water-cooled nuclear reactors combined with water electrolysis compared to the proposed plant design described herein. There is a great deal of interest worldwide in reducing dependence on fossil fuels, while also minimizing the impact of the energy sector on global climate change. One potential opportunity to contribute to this effort is to replace the use of fossil fuels for hydrogen production by the use of water-splitting powered by nuclear energy. Hydrogen production is required for fertilizer (e.g. ammonia) production, oil refining, synfuels production, and other important industrial applications. It is typically produced by reacting natural gas, naphtha or coal with steam, which consumes significant amounts of energy and produces carbon dioxide as a byproduct. In the future, hydrogen could also be used as a transportation fuel, replacing petroleum. New processes are being developed that would permit hydrogen to be produced from water using only heat or a combination of heat and electricity produced by advanced, high temperature nuclear reactors. The U.S. Department of Energy (DOE) is developing these processes under a program known as the Nuclear Hydrogen Initiative (NHI). The Republic of South Africa (RSA) also is interested in developing advanced high temperature nuclear reactors and related chemical processes that could produce hydrogen fuel via water-splitting. This report focuses on the analysis of a nuclear hydrogen production system that combines the Pebble Bed Modular Reactor (PBMR), under development by

  11. Pengolahan Limbah Rumah Makan dengan Proses Biofilter Aerobik

    Directory of Open Access Journals (Sweden)

    Laily Zoraya Zahra

    2015-03-01

    Full Text Available Berkembangnya rumah makan/restoran yang semakin pesat dapat dipastikan akan turut menambah buangan air limbah domestik dengan kadar organik yang tinggi dalam jumlah yang tidak sedikit yang dibuang ke badan air.Tingginya kadar organik dalam limbah domestik rumah makan akan menyebabkan aroma yang tidak sedap jika tidak ada pengolahan terlebih dahulu, maka pengolahan yang dapat digunakan untuk mengolah air limbah rumah makan tersebut adalah proses biofilter aerobik. Penelitian dilakukan dengan proses biofilter aerobik dengan aliran downflow dan menggunakan sistem intermitten.Variabel dalam penelitian ini adalah media biofilter berupa kerikil dan batu alam serta Hydraulic Retention Time (HRT 8 jam. Parameter pencemar yang diukur efisiensinya adalah BOD, COD, dan TSS. Besarnya penyisihan parameter BOD, COD dan TSS dengan menggunakan biofilter aerob berturut-turut mencapai 94,83%, 92,95%, dan 95%. Reaktor paling baik dalam mengolah air limbah adalah reaktor biofilter dengan media kerikil pada HRT 8 jam.

  12. Consequence Analysis of the MHTGR and PBMR

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Yang, Joon Eon; Lee, Won Jae

    2006-01-01

    The probabilistic safety assessment of the VHTR design provides a systematic analysis to identify and quantify all risks that the plant imposes to the operators, general public, and the environment and thus demonstrates compliance to regulatory risk criteria. During the preliminary conceptual design of VHTR in Korea, both block- and pebble type-fuel are considered. Therefore, the consequence analysis of VHTR using both types of fuel were made in order to obtain the basic insights for the classification of events and the formation of the PSA framework of the VHTR

  13. RA reactor operation and maintenance in 1992, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktoro RA u 1992. Godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Tanaskovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-12-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems. [Serbo-Croat] U toku 1992 godine poslovi u okviru projekta 'Istrazivacki nuklearni reaktor RA' obavljani su u skladu sa programom i planom rada. Osnovne aktivnosti na kojima je radjeno odnosile su se na revitalizaciju reaktora RA, kao i na odrzavanje opreme. U ovom periodu reaktor nije bio u pogonu. Svo osoblje je bilo angazovano na poslovima rekonstrukcije i modernizacije postojecih i dogradnje novih reaktorskih sistema, na odrzavanju opreme a deo tehnickog osoblja je bio obucavan za vrsenje odgovarajucih poslova u pogonu i odrzavanju opreme.

  14. Activities of RAHP and problems to be cleared towards commercialization

    International Nuclear Information System (INIS)

    An, S.; Tsuchie, Y.

    2001-01-01

    Research Association of HTGR Plant (RAHP) is the sole research association on HTGR Plants in industrial sector of Japan since its establishment in 1985. Activities in these years were to analyze world R and D trends, to set-up fundamental development strategies, and to incorporate the study results into actions towards commercialization of HTGR. RAHP is now mainly investigating fuel cycle aspects and development programs of small modular reactors, such as US-Russian GT-MHR and South African PBMR. Conclusions obtained through the activities so far are: (a) From view points of effective use of energies and reduction of environmental impacts in global scale, development of nuclear power is essential, and that of HTGR is desirable in particular, because of its very highly inherent safety and feasibility of high temperature heat utilizations. The role of HTGR is inter-complementary with those of LWR and FBR. (b) Future subjects and problems to be cleared towards commercializing HTGR are: (1) safety, technical and economical demonstrations, including inherent safety, fuel integrity as key of the safety, gas turbine cycle, high temperature system performances, etc.; (2) acceptance of safety design concept, and preparation for its evaluation and licensing. (c) Construction of demonstration plant(s) as a total system, through effective international cooperation, such as internationally common or sharing development on basic machinery systems between GT-MHR and PBMR Programs, becomes most significant. (author)

  15. Enhancing the Sustainability of Nuclear Power: the Pebble Bed HTR in Deep Burn Mode

    International Nuclear Information System (INIS)

    Da Cruz, D.F.; De Haas, J.B.M.; Van Heek, A.I.

    2004-01-01

    The scenario of a utility in an industrialized country starting new nuclear construction with a single PBMR reactor has been considered. To make the new construction project acceptable by government and society, a maximum effort to obtain sustainability (i.e. minimization of resource use and waste production) will have to be shown. Therefore the usual open cycle for HTR has been abandoned, and the spent fuel will be reprocessed once. The long-lived transuranic (TRU) elements Pu, Np, Am and Cm are all re-fabricated into so-called transmutation fuel elements, and loaded back into the same reactor, in our case a 110 MWe PBMR with low-enriched uranium cycle. In this study, the reactor physical prospects have been investigated: to what extent the amount of TRU could be reduced. In this way, 75% of the initial amount of TRU waste is being destructed, while the time span in which the waste is more radio-toxic than uranium ore is being reduced to one-third. Also, the amount of fresh driver fuel needed is decreases by 25%. A preliminary cost analysis has been performed as well. It shows that there is also a cost advantage of operating the reactor in Deep Burn mode in industrialized countries, where the waste storage fees charged per volume are relatively high. (authors)

  16. Steam process cogeneration using nuclear energy

    International Nuclear Information System (INIS)

    Alonso, G.; Ramirez, R.

    2010-10-01

    Use of energy in a sustainable manner is to make processes more efficient. Oil industry requires of electricity and steam for refinery and petrochemical processes, nuclear energy can be a clean energy alternative. Cogeneration is an option to be assessed by Mexico to provide additional value to electricity generation. Mexico is a country with oil resources that requires process heat for gasoline production among other things. With the concern about the climate change and sustain ability policies it is adequate to use cogeneration as a way to optimize energy resources. Currently there is a national program that considers cogeneration for several Mexican refineries, and the first choices are combined cycle plants and thermo power plants using residual oil. This is long term program. The pebble bed modular reactor (PBMR) is a next generation reactors that works with very high temperatures that can be used to produce steam process along with electricity, in this work two different couplings are assessed for the PBMR reactor to produce steam process, the two couplings are compared for using in the Mexican refineries and some conclusions are given. (Author)

  17. Process heat cogeneration using a high temperature reactor

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Ramirez, Ramon; Valle, Edmundo del; Castillo, Rogelio

    2014-01-01

    Highlights: • HTR feasibility for process heat cogeneration is assessed. • A cogeneration coupling for HTR is proposed and process heat cost is evaluated. • A CCGT process heat cogeneration set up is also assessed. • Technical comparison between both sources of cogeneration is performed. • Economical competitiveness of the HTR for process heat cogeneration is analyzed. - Abstract: High temperature nuclear reactors offer the possibility to generate process heat that could be used in the oil industry, particularly in refineries for gasoline production. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product and if the cost of this subproduct will be competitive with other alternatives. The current study assesses the likeliness of generating process heat from Pebble Bed Modular Reactor to be used for a refinery showing different plant balances and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor where the cycle configuration to transport the heat of the reactor to the process plant plays an important role in the cycle efficiency and in the plant economics. The results of this study show that the PBMR would be most competitive when capital discount rates are low (5%), carbon prices are high (>30 US$/ton), and competing natural gas prices are at least 8 US$/mmBTU

  18. Process heat cogeneration using a high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Gustavo, E-mail: gustavoalonso3@gmail.com [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico); Instituto Politécnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ramirez, Ramon [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico); Valle, Edmundo del [Instituto Politécnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Castillo, Rogelio [Instituto Nacional de Investigaciones Nucleares, Carretera Mexico-Toluca s/n, Ocoyoacac, Edo. De Mexico 52750 (Mexico)

    2014-12-15

    Highlights: • HTR feasibility for process heat cogeneration is assessed. • A cogeneration coupling for HTR is proposed and process heat cost is evaluated. • A CCGT process heat cogeneration set up is also assessed. • Technical comparison between both sources of cogeneration is performed. • Economical competitiveness of the HTR for process heat cogeneration is analyzed. - Abstract: High temperature nuclear reactors offer the possibility to generate process heat that could be used in the oil industry, particularly in refineries for gasoline production. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product and if the cost of this subproduct will be competitive with other alternatives. The current study assesses the likeliness of generating process heat from Pebble Bed Modular Reactor to be used for a refinery showing different plant balances and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor where the cycle configuration to transport the heat of the reactor to the process plant plays an important role in the cycle efficiency and in the plant economics. The results of this study show that the PBMR would be most competitive when capital discount rates are low (5%), carbon prices are high (>30 US$/ton), and competing natural gas prices are at least 8 US$/mmBTU.

  19. Karakterisasi Unjuk Kerja Diesel Engine Generator Set Sistem Dual Fuel Solar-Syngas Hasil Gasifikasi Briket Municipal Solid Waste (MSW Secara Langsung

    Directory of Open Access Journals (Sweden)

    Achmad Rizkal

    2017-01-01

    Full Text Available Sejalan dengan semakin banyaknya kebutuhan energi untuk dapat digunakan sebagai bahan bakar maka perlu adanya pengembangan gas biomassa sebagai bahan bakar alternatif pada motor pembakaran dalam maka akan dilakukan penelitian mengenai aplikasi sistem dual fuel gas hasil gasifikasi biomassa municipal solid waste (msw pada sistem downdraft dengan minyak solar pada motor diesel stasioner. Penelitian ini bertujuan untuk mengetahui seberapa besar solar yang tersibtitusi dengan adanya penambahan syngas yang disalurkan secara langsung. Penelitian ini dilakukan secara eksperimental dengan proses pemasukan aliran syngas yang dihasilkan downdraft municipal solid waste (MSW kedalam saluran udara mesin diesel generator set secara langsung menggunakan sistem mixer. Pengujian dilakukan dengan putaran konstan 2000 rpm dengan pembebanan bervariasi dari 200 watt sampai dengan 2000 watt dengan interval 200 watt. Bahwa produksi syngas dari reaktor gasifikasi ditambahkan sistem bypass untuk mengetahui kesesuaian antara reaktor gasifikasi dan mesin generatorset data ṁ syngas yang dibutuhkan mesin diesel, ṁ syngas yang di bypass untuk mendapatkan kesesuaian antara produksi syngas dan yang di bypass.  Data-data yang diukur dari penelitian ini menunjukkan bahwa besar nilai mass flowrate gas syngas yang dibutuhkan mesin diesel pada AFR reaktor gasifier 1,39 sebesar 0,0003748 kg/s. Mass flowrate gas syngas yang di bypass menunjukkan nilai 0 pada saat sistem dijalankan karena seluruh gas syngas masuk kedalam ruang bakar. AFR rata-rata sebesar 14,54 ,Nilai Spesifik fuel consumption (sfc mengalami peningkatan 68% dari kondisi standar single fuel , Nilai efesiensi thermal mengalami kenaikan sebesar 7% dari kondisi single fuel, Nilai daya rata-rata sebesar 2,28kW, Nilai torsi rata-rata sebesar 10,94 N.m. Solar yang tersibtitusi sebesar 48%. Nilai temperatur (coolant, mesin, oil, dan gas buang pada setiap pembebanan mengalami kenaikan.

  20. ANALISA PENGASUTAN MOTOR INDUKSI 3 FASA 2500 KW SEBAGAI PENGGERAK FAN PADA BAG FILTER

    Directory of Open Access Journals (Sweden)

    Budi Yanto Husodo

    2017-11-01

    Full Text Available Salah satu persoalan yang timbul pada pengoperasian motor induksi adalah arus pengasutan yang tinggi yang nilainya bisa mencapai sepuluh kali arus nominal. Arus pengasutan yang besar ini mengakibatkan penurunan tegangan sesaat (sag pada sistem jaringan. Selain itu juga menyebabkan tingginya pemakaian daya hingga sebesar 1,5 - 2,5 kali daya nominal yang berakibat pada tingginya energi pemakaian pada saat pengasutan. Metode pengasutan diperlukan untuk mengurangi arus pengasutan dan pemakaian energi yang besar tersebut. Pada makalah ini dilakukan perbandingan tiga metode pengasutan motor induksi yaitu berupa autotrafo, reaktor dan rangkaian star-delta. Dengan menggunakan software ETAP, pengujian dilakukan pada motor induksi 6 kV, 279 A, 2500 kW, dan faktor daya sebesar 0,879, sebagai penggerak fan pada bag filter. Hasil simulasi menunjukkan bahwa autotrafo memberikan penurunan arus pengasutan yang paling besar yaitu 73,82% dari arus pengasutan motor tanpa bantuan alat pengasutan. Sedangkan konsumsi energi yang paling kecil didapatkan dengan mengunakan pengasutan reaktor, di mana energi pemakaian berkurang dari 31,102 kWh tanpa pengasutan menjadi 17,676 kWh atau setara dengan 43,17%.

  1. South African safety assessment framework for the pebble bed modular reactor - HTR2008-58192

    International Nuclear Information System (INIS)

    Joubert, J.; Kohtz, N.; Coe, I.

    2008-01-01

    It is planned to construct a first of a kind Pebble Bed Modular Reactor (PBMR) in South Africa. A need has been recognized to accompany the licensing process for the PBMR with independent safety assessments to ensure that the safety case submitted by the applicant complies with the licensing requirements of the NNR. At the HTR 2006 Conference, the framework and major challenges on safety assessment that the South African National Nuclear Regulator (NNR) faces in developing and applying appropriate strategies and tools were presented. This paper discusses the current status of the various NNR assessment activities and describes how this will be considered in the NNR Final Report on the PBMR Safety Case. The traditional safety assessment process has been adapted to take into account the developmental nature of the project. By performing safety assessments, the designer and applicant must ensure that the design as proposed for construction and as-built meets the safety requirements defined by the regulatory framework. The regulator performs independent safety assessments, including independent analyses in areas deemed safety significant and potentially safety significant. The developmental nature of the project also led to the identification of a series of regulatory assessment activities preceding the formal assessment of the safety case. Besides an assessment of the resolution of Key Licensing Issues which have been defined in an early stage of the project and are discussed in /l/, these activities comprise the participation in an SAR Early Intervention Process, the execution of a regulatory HAZOP and the development of a regulatory assessment specification for the formal assessment of the safety case. This paper briefly describes these activities and their current status. During the last two years, significant progress was made with the development or adjustment of tools for the independent analysis by the regulator of the steady state core design, of the transient

  2. The influence of thorium on the temperature reactivity coefficient in a 400 MWth pebble bed high temperature plutonium incinerating reactor

    International Nuclear Information System (INIS)

    Richards, Guy A.; Serfontein, Dawid E.

    2014-01-01

    This article investigates advanced fuel cycles containing thorium and reactor grade plutonium (Pu(PWR)) in a 400 MW th Pebble Bed Modular Reactor (PBMR) Demonstration Power Plant. Results presented were determined from coupled neutronics and thermo-hydraulic simulations of the VSOP 99/05 diffusion codes. In a previous study impressive burn-ups (601 MWd/kg heavy metal (HM)) and thus plutonium destruction rates (69.2 %) were obtained with pure plutonium fuel with mass loadings of 3 g Pu(PWR)/fuel sphere or less. However the safety performance was poor in that the limit on the maximum fuel temperature during equilibrium operation was exceeded and positive Uniform Temperature Reactivity Coefficients (UTCs) were obtained. In the present study fuel cycles containing mixtures of thorium and plutonium achieved negative maximum UTCs. Plutonium only fuel cycles also achieved negative maximum UTCs, provided that much higher mass loadings are used. It is proposed that the lower thermal neutron flux was responsible for this effect. The plutonium only fuel cycle with 12 g Pu(PWR)/fuel sphere also achieved the adopted safety limits for the PBMR DPP-400 in that the maximum fuel temperature and the maximum power density did not exceed 1130°C or 4.5 kW/sphere respectively. This design would thus be licensable and could potentially be economically feasible. However the burn-up was much lower at 181 MWd/kgHM and thus the plutonium destruction fraction was also much lower at 24.5%, which may be sub-optimal with respect to proliferation and waste disposal objectives and therefore further optimisation studies are proposed. (author)

  3. Ultrasonics aids the identification of failed fuel rods

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Over a number of years Brown Boveri Reaktor of West Germany has developed and commercialized an ultrasonic failed fuel rod detection system. Sipping has up to now been the standard technique for failed fuel detection, but sipping can only indicate whether or not an assembly contains defective rods; the BBR system can tell which rod is defective. (author)

  4. OTTO-PAP: An alternative option to the PBMR fuelling philosophy

    International Nuclear Information System (INIS)

    Mulder, E.; Teuchert, E.

    1997-01-01

    Once Through Then Out, Power Adjusted by Poison (OTTO-PAP) fuelling of a high temperature pebble-bed reactor offers a simple alternative to the MEDUL (Mehrfachdurchlauf = German for multi-pass) fuelling regime followed in pebble bed reactor designs to date. The prerequisite for a modular reactor unit of maximum power output, subject to observing passive safety characteristics is a sufficiently flat axial neutron flux profile. This is achieved by introducing B 4 C coated particles of pre-calculated size and packing density within the fuel spheres. In accordance with AVR operating practise the temperature profile is radially equalised by introducing a 2-zone core loading. Adding pure graphite spheres loosely into the centre column area of the core effectively reduced the maximum power in the middle. Increasing the reactor diameter is enabled by the introduction of noses. A 3-D geometric modeller developed in cylindrical co-ordinates enables a given flow description of the pebbles adjacent to the nose boundaries and in the vicinity of the shut down/control rods. After translation of the geometric data the neutronic behaviour of the reactor is followed in 3-D by the CITATION code. This study is aimed towards achieving an optimal core layout with a LEU (Low Enriched Uranium) fuel cycle. Physical properties of the OTTO-PAP, 150 MWt reference design is reported, while computations performed observe results obtained by the reference HTR-MODUL design. (author)

  5. RA reactor operation and maintenance in 1994, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1994. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Mikic, N; Tanaskovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1994-12-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Zavrseni su radovi na izgradnji sistema za udesno hladjenje, rekonstrukciji postojeceg sistema specijalne ventilacije i rekonstrukciji sistema za napajanje elektricnom energijom. Zapoceti su radovi na modernizaciji, odnosno zameni instrumentacije reaktora ali njegova realizacija kasni

  6. PERANCANGAN PEMBANGKIT LISTRIK TENAGA SAMPAH ORGANIK ZERO WASTE DI KABUPATEN TEGAL (STUDI KASUS DI TPA PENUJAH KABUPATEN TEGAL

    Directory of Open Access Journals (Sweden)

    Abdul Muiz Liddinillah Sanfiyan

    2017-12-01

    Full Text Available Permasalahan sistem pengolahan sampah yang ada di Kabupaten Tegal adalah masih menggunakan sistem Open Dumping. Berdasarkan data yang diperoleh dari Badan Pusat Statistik (BPS Kabupaten Tegal pada tahun 2016, komposisi sampah organik adalah yang terbesar kedua setelah sampah plastik dan sangat berpotensi mengalami penambahan setiap tahunnya. Tujuan dari penelitian ini adalah untuk membuat perancangan pembangkit listrik tenaga sampah organik zero waste di Kabupaten Tegal, dengan studi kasus di Tempat Pembuangan Akhir (TPA sampah Penujah. Objek dalam penelitian ini adalah sistem pengolahan sampah organik yang ada di Kabupaten Tegal dengan menggunakan sistem pengolahan sampah zero waste. Sistem pengolahan sampah organik zero waste adalah sistem pengolahan sampah yang tidak menghasilkan sampah kembali. Jadi, diharapkan jumlah sampah organik akan berkurang secara bertahap. Sampah organik dapat dirubah menjadi biogas melalui proses fermentasi yang dibantu oleh bakteri secara anaerob di dalam reaktor biodigester. Biogas tersebut ditampung di dalam tempat penampungan untuk kemudian didistribusikan ke dalam genset biogas sebagai bahan bakar pembangkit listrik. Sisa pengolahan biogas dapat dirubah menjadi pupuk cair dan pupuk kompos yang bernilai ekonomis.

  7. Challenges in developing human resources for nuclear safety in South Africa

    Energy Technology Data Exchange (ETDEWEB)

    Tsatsi, Louisa [National Nuclear Regulator, PO BOX 7106, 00046 Centurion (South Africa)

    2008-07-01

    Challenges in developing Human Resources for nuclear safety in South Africa ESKOM Holding Limited which is the South African Government owned utility, operates over 10 power stations. The total installed is about 40 GW, and nuclear contributes only 6 percent. The existing nuclear power station, Koeberg NPP, is comprised of two 900 MW(e) units at the South African west coast near Cape Town. The South African Government has a policy to increase the share of nuclear in the generation mix from 6 percent to 15 percent before the year 2020. The challenge is that there have been a 'greying' of nuclear experts and a shrinking of nuclear engineering and science departments. As a consequence of this, ESKOM has realized that a large number of young engineers and scientists would have to be recruited and then trained in South Africa and abroad. Some people, especially high performers in the industry are continually looking for new challenges and opportunities and though it is important in the nuclear industry to retain these key staff members it have proved to be a serious challenge. The nuclear industry had to consider their national training infrastructures and the South African government in partnership with ESKOM, NECSA and PBMR has started a process of funding university chairs in reactor engineering and allied subjects. These departments undertake research and provide training for the South African nuclear industry. The recruitment process has initially involved the transfer of personnel from ESKOM, NECSA as well as direct recruitment from the market. The primary recruitment process going forward will be from universities and other Further Education Training (FET) institutions with a focus on ESKOM and PBMR providing the specific nuclear training. In this regard, both ESKOM and PBMR provide bursaries, project work and other assistance to selected candidates. Upon completion of studies, the specific training is provided both in-house and with partner national

  8. Challenges in developing human resources for nuclear safety in South Africa

    International Nuclear Information System (INIS)

    Tsatsi, Louisa

    2008-01-01

    Challenges in developing Human Resources for nuclear safety in South Africa ESKOM Holding Limited which is the South African Government owned utility, operates over 10 power stations. The total installed is about 40 GW, and nuclear contributes only 6 percent. The existing nuclear power station, Koeberg NPP, is comprised of two 900 MW(e) units at the South African west coast near Cape Town. The South African Government has a policy to increase the share of nuclear in the generation mix from 6 percent to 15 percent before the year 2020. The challenge is that there have been a 'greying' of nuclear experts and a shrinking of nuclear engineering and science departments. As a consequence of this, ESKOM has realized that a large number of young engineers and scientists would have to be recruited and then trained in South Africa and abroad. Some people, especially high performers in the industry are continually looking for new challenges and opportunities and though it is important in the nuclear industry to retain these key staff members it have proved to be a serious challenge. The nuclear industry had to consider their national training infrastructures and the South African government in partnership with ESKOM, NECSA and PBMR has started a process of funding university chairs in reactor engineering and allied subjects. These departments undertake research and provide training for the South African nuclear industry. The recruitment process has initially involved the transfer of personnel from ESKOM, NECSA as well as direct recruitment from the market. The primary recruitment process going forward will be from universities and other Further Education Training (FET) institutions with a focus on ESKOM and PBMR providing the specific nuclear training. In this regard, both ESKOM and PBMR provide bursaries, project work and other assistance to selected candidates. Upon completion of studies, the specific training is provided both in-house and with partner national and

  9. Services provided to other organizational units, Annex 7; Prilog 7 - Usluge drugim organizacionim jedinicama

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1965-12-15

    During 1965, the RA reactor staff provided services to other organizational units in: designing new systems, repair of electronic equipment, installing measuring devices, constructing mechanical elements in the workshop. [Serbo-Croat] U toku 1965. godine reaktor RA je pruzio usluge drugim jedinicama i to: projektovanjem novih sistema, popravkom elektronskih uredjaja, instaliranjem mernih sistema, izradom mehanickih elemenata i sklopova u mehanickoj radionici.

  10. Servicing NPPs in Japan, Korea and Taiwan

    International Nuclear Information System (INIS)

    Bohmann, W.; Poetz, F.

    1991-01-01

    More than 45 comparable orders (for equipment or services ) have been carried out in Japan, Korea and Taiwan by ABB Reaktor since 1982. Recently contracts to deliver inspection and repair equipment for the core baffle former bolts of Japanese NPPs, and in 1990, a contract to clean two steam generators in Korea, together with Pacific Nuclear Services, USA, were won. 2 figs

  11. INVESTIGATION ON THERMAL-FLOW CHARACTERISTICS OF HTGR CORE USING THERMIX-KONVEK MODULE AND VSOP'94 CODE

    Directory of Open Access Journals (Sweden)

    Sudarmono Sudarmono

    2015-03-01

      Kegagalan sistem pembuangan panas pada reaktor berpendingin air jenis PWR, Three Mile Islands dan reaktor BWR Fukushima Daiichi, menyebabkan masyarakat nuklir mulai memikirkan penggunaan reaktor pembangkit daya jenis temperatur tinggi berpendingin gas (HTGR. Bidang Fisika dan Teknologi Reaktor di Pusat Teknologi Reaktor dan Keselamatan Nuklir (PTRKN mempunyai tugas melaksanakan kegiatan litbang desain konseptual reaktor kogenerasi dengan tingkat daya menengah yang berpendingin gas helium dengan daya 200 MWt. Desain HTGR200K merupakan salah satu sistem pembangkit energi yang memiliki efisiensi energi paling besar, dan tingkat keselamatan inheren yang tinggi dan bersih. Komposisi geometri dan struktur teras didesain agar dapat menghasilkan keluaran pendingin gas helium bertemperatur 950 0C sehingga dapat digunakan untuk produksi hidrogen dan atau unit industri proses lainnya secara kogeneratif. Luaran gas helium bertemperatur sangat tinggi ini akan menimbulkan tegangan termal pada bola bahan bakar yang mengancam integritas sistem pengungkungan produk fisi di dalamnya. Oleh karena itu perlu dilakukan evaluasi karakteristika termal flow untuk menentukan distribusi temperatur bahan bakar bola dan outlet temperatur pendingin gas helium teras HTGR. Hal ini dilakukan dengan menggunakan modul Thermix-Konvek yang terintegrasi dalam program VSOP’94. Geometri teras HTGR dikerjakan dalam modul BIRGIT untuk model teras 2-D (R-Z dengan 5 kanal aliran pebble dalam teras aktif arah radial. Hasil evaluasi menunjukkan bahwa nilai tertinggi dan terendah temperatur yang terdapat pada teras   adalah sebesar 999.3 °C dan 886,5 °C. Demikian pula hasil temperatur tertinggi bahan bakar TRISO dan bahan bakar pebble di dalam teras, yaitu diperoleh sebesar  1510,20°C yang terletak pada lapisan bahan bakar inti UO2, di posisi z= 335.51 cm dan  r=0 cm. Analysis di lakukan pada laju massa aliran pendingin, tekanan dan daya masing-masing sebesar 120 kg/s, 7 Mpa dan 200MWth. Hasil

  12. ANALISA PENGARUH SUHU AWAL PELAT PANAS PADA PROSES QUENCHING CELAH SEMPIT REKTANGULAR

    Directory of Open Access Journals (Sweden)

    M. Hadi Kusuma

    2015-03-01

    Full Text Available Pemahaman terhadap manajemen termal apabila terjadi suatu kecelakaan parah reaktor nuklir seperti melelehnya bahan bakar dan teras reaktor, menjadi prioritas utama untuk menjaga integritas bejana tekan reaktor. Dengan demikian hasil lelehan bahan bakar dan teras reaktor (debris tidak keluar dari bejana tekan reaktor dan mengakibatkan dampak lain yang lebih besar ke lingkungan. Salah satu cara yang dilakukan untuk menjaga integritas bejana tekan reaktor adalah dengan melakukan pendinginan terhadap panas berlebih yang dihasilkan akibat dari kecelakaan tersebut. Untuk mempelajari dan mendapatkan pemahaman mengenai hal tersebut, maka dilakukan penelitian mengenai pengaruh suhu awal pelat panas dalam proses quenching (pendinginan secara tiba-tiba celah sempit rektangular. Penelitian difokuskan pada penentuan suhu rewetting dari pendinginan pelat panas dengan suhu awal pelat 220 0C, 400 0C, dan 600 0C dengan laju aliran air pendingin 0,2 liter/detik. Eksperimen dilakukan dengan menginjeksikan air pada laju aliran 0,2 liter/detik pada suhu air pendingin 85 0C ke dalam celah sempit rektangular. Data hasil pengukuran digunakan untuk mengetahui suhu rewetting yang terjadi pada pendinginan pelat panas tersebut. Tujuannya adalah untuk memahami pengaruh suhu awal pelat panas terhadap rewetting pada proses quenching di celah sempit rektangular. Hasil yang diperoleh menunjukkan bahwa titik rewetting pada pendinginan pelat panas 220 0C, 400 0C, dan 600 0C terjadi pada suhu rewetting yang berbeda-beda. Pada suhu awal pelat panas 220 0C, suhu rewetting terjadi pada 220 0C yaitu langsung ketika air dilewatkan melalui celah sempit rektangular. Pada suhu awal pelat panas 400 0C, suhu rewetting terjadi pada 379,51 0C. Dan pada suhu awal pelat panas 600 0C, suhu rewetting terjadi pada 426,63 0C. Perbedaan suhu awal pelat panas yang sangat signifikan menyebabkan terjadinya perubahan sifat fisik benda uji, berbedanya rejim pendidihan yang dialami oleh fluida yang

  13. Synergistic use of active and passive microwave in soil moisture estimation

    Science.gov (United States)

    O'Neill, P.; Chauhan, N.; Jackson, T.; Saatchi, S.

    1992-01-01

    Data gathered during the MACHYDRO experiment in central Pennsylvania in July 1990 have been utilized to study the synergistic use of active and passive microwave systems for estimating soil moisture. These data sets were obtained during an eleven-day period with NASA's Airborne Synthetic Aperture Radar (AIRSAR) and Push-Broom Microwave Radiometer (PBMR) over an instrumented watershed which included agricultural fields with a number of different crop covers. Simultaneous ground truth measurements were also made in order to characterize the state of vegetation and soil moisture under a variety of meteorological conditions. A combination algorithm is presented as applied to a representative corn field in the MACHYDRO watershed.

  14. Technical status of the pebble bed modular reactor (PBMR-SA) conceptual design

    International Nuclear Information System (INIS)

    Fox, M.

    1997-01-01

    The reactor study is well underway seen from a broad spectrum of disciplines and technology. The objective power output with a high efficiency direct cycle power conversion unit remains promising after compiling the first critical analysis of the core and the power conversion unit. The stability and controllability of the system are demonstrated by the engineering simulator. The main system and components are basically specified for costing purposes. A first plant layout has been completed demonstrating the positions of main components, personnel movement, installation methods for large components, etc. A cryptic report style presentation includes study objectives, indicating guiding documents, giving an overview of design and analyses work done as well as a few sketches and diagram are included in this paper. Most of these sketches and diagrams are small replicas of large drawings and are therefore not readable but can be used as references. (author)

  15. Data concerning operation and application of RA reactor in 1975, Annex 1; Prilog 1 - Podaci o radu i iskoriscenosti reaktora RA u 1975. godini

    Energy Technology Data Exchange (ETDEWEB)

    Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-01-15

    RA reactor was operating according to the plan for 1975 adopted in December of the previous year. It was planned for reactor to be operated at nominal power first 10-20 days each month, three following days were reserved for different power levels according to the users' demand. Four fuel exchanges were planned and fulfilled with minor delay. Data concerning planned and real operation, as well as delays from the plan and shorter interruptions are presented in tables of this Annex. It is shown that all the delays and interruptions which amounted to 104 hours were compensated. [Serbo-Croat] Reaktor RA je radio prema planu rada za 1975. godinu, nacinjenom u decembru prethodne godine. Planirano je da reaktor radi neprekidno prvih 10-20 dana u mesecu na nominalnoj snazi, tri sledeca dana je rezervisano za rad na drugim snagama zavisno od potreba korisnuka. Planirane su i 4 izmene goriva. Podaci o planiranom i ostvarenom radu kao i odstupanjima od plana i kracim prekidima u radu reaktora dati su u tabelama ovog priloga. Vidi se da su sva odstupanja i prekidi u ukupnom trajanju od 104 sata u celini nadoknadjeni.

  16. Quality management in an international nuclear power plant project

    International Nuclear Information System (INIS)

    Brion, J.; Crustin, J.

    1975-01-01

    SNR (Schneller Natriumgekuehlter Reaktor) is the fast reactor power plant being erected at Kalkar, Federal German Republic. Quality management in this project is a contractual obligation. Quality management is subdivided into quality engineering, set of actions performed before manufacturing, and quality control, set of material controls performed during fabrication. The two successive phases of the quality management are presented. The difficulties and improvment possibilities are discussed [fr

  17. The reactor accident in the Three Mile Island-2 reactor plant. (Harrisburg, USA)

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The presentation of the accident development is based on the bulletin of the Atomic Industrial Forum (AIF) from the 6th of April, 1979. In addition, there are some short pieces of information, No. 14 and 15 of the association for reactor security as well as written and verbal information of the firms Brown, Boveri and Co/Babcock-Brown Boveri Reaktor GmbH (BBC/BBR). (orig.) [de

  18. NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2016-03-01

    Full Text Available Abstract NEUTRONICS ANALYSIS ON MINI TEST FUEL IN THE RSG-GAS CORE. Research of UMo fuel for research reactor has been developing  right now. The fuel of  research reactor used is uranium low enrichment with high density. For supporting the development of fuel, an assessment of mini fuel in the RSG-GAS core was performed. The mini fuel are U7Mo-Al and U6Zr-Al with densitis of 7.0gU/cc and 5.2 gU/cc, respectively. The size of both fuel are the same namely 630x70.75x1.30 mm were inserted to the 3 plates of dummy fuel. Before being irradiated in the core, a calculation for safety analysis  from neutronics and thermohydrolics aspects were required. However, in this paper will discuss safety analysis of the U7Mo-Al and U6Zr-Al mini fuels from neutronic point of view.  The calculation was done using WIMSD-5B and Batan-3DIFF code. The result showed that both of the mini fuels could be irradiated in the RSG-GAS core with burn up less than 70 % within 12 cycles of operation without over limiting the safety margin. Power density of U7Mo-Al mini fuel bigger than U6Zr-Al fuel.   Key words: mini fuel, neutronics analysis, reactor core, safety analysis   Abstrak ANALISIS NEUTRONIK ELEMEN BAKAR UJI MINI DI TERAS RSG-GAS. Penelitian tentang bahan bakar UMo untuk reaktor riset terus berkembang saat ini. Bahan bakar reaktor riset yang digunakan adalah uranium pengkayaan rendah namun densitas tinggi.  Untuk mendukung pengembangan bahan bakar dilakukan uji elemen bakar mini di teras reakror RSG-GAS dengan tujuan menentukan jumlah siklus di dalam teras sehingga tercapai fraksi bakar maksimum. Bahan bakar yang diuji adalah U7Mo-Al dengan densitas 7,0 gU/cc dan U6Zr-Al densitas 5,2 gU/cc. Ukuran kedua bahan bakar uji tersebut adalah sama 630x70,75x1,30 mm dimasukkan masing masing kedalam 3 pelat dummy bahan bakar. Sebelum diiradiasi ke dalam teras reaktor maka perlu dilakukan perhitungan keselamatan baik secara neutronik maupun termohidrolik. Dalam makalah ini

  19. DESAIN KONSEP TANGKI PENAMPUNG BAHAN BAKAR PASSIVE COMPACT MOLTEN SALT REACTOR

    Directory of Open Access Journals (Sweden)

    A. Hadiwinata

    2015-04-01

    Full Text Available Passive Compact Molten Salt Reactor (PCMSR merupakan pengembangan dari reaktor MSR. Desain reaktor PCMSR membutuhkan tempat khusus penampung sementara bahan bakar pada saat terjadi insiden, misalnya kecelakaan yang menyebabkan peningkatan suhu bahan bakar. Tangki penampung bahan bakar tersusun dari 3 bagian yang saling terhubung yaitu bagian penampung cairan bahan bakar, cerobong (chimney, dan penukar kalor. Dalam penelitian ini, tangki dimodelkan secara lump dan dilakukan variasi daya awal reaktor dan ketinggian cerobong. Syarat batas model ditetapkan suhu bahan bakar maksimum 1400 °C, yang didasarkan pada titik didih larutan garam LiF-BeF2-ThF4-UF4. Analisis dilakukan dengan cara menghitung rugi tekanan total dan transfer kalor untuk variasi daya awal antara 1800-3000 MWth dan ketinggian cerobong antara 1-10 m. Hasil penelitian menunjukan semakin besar daya reaktor, maka tinggi tangki penampung bahan bakar dan tinggi alat penukar kalor yang dibutuhkan akan semakin besar, tejadi kenaikan suhu fluida pendingin dan suhu udara pendingin, dan menyebabkan kenaikan laju aliran masa fluida pendingin, sedangkan laju aliran masa udara menurun. Peningkatan ketinggian cerobong menyebabkan ketinggian tangki penampung bahan bakar dan ketinggian alat penukar kalor semakin menurun, penurunan suhu fluida pendingin, tetapi suhu udara meningkat, dan menyebabkan peningkatan laju aliran masa fluida pendingin, tetapi laju aliran masa udara akan semakin menurun. Kata kunci: PCMSR, cerobong, alat penukar kalor, variasi daya.   The Passsive Compact Molten Salat Reactor (PCMSR reactor is developed from MSR reactor. The PCMSR reactor design requires special place to temporarily storage for reactor fuel when incident occurs, such as when there is an accident which caused the temperature of the fuel increases. The tank consist of three interconnected parts, the reservoir liquid fuel, chimney, and the heat exchanger. In this research, the tank system is modeled based on

  20. Hydrogen Mixing Studies (HMS) assessment manual

    International Nuclear Information System (INIS)

    Lam, K.L.; Wilson, T.L.; Travis, J.R.

    1993-06-01

    This report documents some calculations performed to assess the Hydrogen Mixing Studies (HMS) code. Results are presented first for some analytical test problems, including laminar flow and mass diffusion. The von Karman vortex street problem and the Sandia FLAME Facility and Heiss Dampf Reaktor (HDR) containment facility test problems are then discussed. For the analytical problems, the code gave results that agree exceptionally well with the analytical solutions. Calculations for the von Karman vortex street problem were performed at selected Reynolds numbers for several obstacle types. The computed flow patterns agree well with experimental observations-specifically the occurrence of a vortex street (double row of vortices) above a critical Reynolds number. Calculations for the von Karman vortex street problem were performed at selected Reynolds numbers for several obstacle types. The computed flow patterns agree well with experimental observations-specifically the occurrence of a vortex street (double row of vortices) above a critical Reynolds number. The last assessment problem involves modeling the experiment T31.5. The experiment was carried out in the HDR containment building, which is a large, multi-compartment facility (11 300 m 3 free volume in 72 compartments). In the experiment, a steam-water mixture was first injected into the containment to simulate a large-break blowdown of a pressure vessel, and then superheated steam was injected that was followed by a release of helium-hydrogen light gas. The calculated results (pressure, temperature, and gas concentrations) agree reasonably well with the experimental data

  1. Performance review: PBMR closed cycle gas turbine power plant

    International Nuclear Information System (INIS)

    Pradeep Kumar, K.N.; Tourlidakis, A.; Pilidis, P.

    2001-01-01

    Helium is considered as one of the ideal working fluid for closed cycle using nuclear heat source due to its low neutron absorption as well as high thermodynamic properties. The commercial viability of the Helium turbo machinery depends on operational success. The past attempts failed due to poor performances manifested in the form of drop in efficiency, inability to reach maximum load, slow response to the transients etc. Radical changes in the basic design were suggested in some instances as possible solutions. A better understanding of the operational performance is necessary for the detailed design of the plant and the control systems. This paper describes the theory behind the off design and transient modelling of a closed cycle gas turbine plant. A computer simulation model has been created specifically for this cycle. The model has been tested for various turbine entry temperatures along the steady state and its replications at various locations were observed. The paper also looks at the various control methods available for a closed cycle and some of the options were simulated. (author)

  2. Pebble bed modular reactor - The first Generation IV reactor to be constructed

    International Nuclear Information System (INIS)

    Ion, S.; Nicholls, D.; Matzie, R.; Matzner, D.

    2004-01-01

    Substantial interest has been generated in advanced reactors over the past few years. This interest is motivated by the view that new nuclear power reactors will be needed to provide low carbon generation of electricity and possibly hydrogen to support the future growth in demand for both of these commodities. Some governments feel that substantially different designs will be needed to satisfy the desires for public perception, improved safety, proliferation resistance, reduced waste and competitive economics. This has motivated the creation of the Generation IV Nuclear Energy Systems programme in which ten countries have agreed on a framework for international cooperation in research for advanced reactors. Six designs have been selected for continued evaluation, with the objective of deployment by 2030. One of these designs is the very high temperature reactor (VHTR), which is a thermal neutron spectrum system with a helium-cooled core utilising carbon-based fuel. The pebble bed modular reactor (PBMR), being developed in South Africa through a worldwide international collaborative effort led by Eskom, the national utility, will represent a key milestone on the way to achievement of the VHTR design objectives, but in the much nearer term. This paper outlines the design objectives, safety approach and design details of the PBMR, which is already at a very advanced stage of development. (author)

  3. Efektivitas Pengolahan Air Limbah Batik dengan Cara Kimia dan Biologi

    Directory of Open Access Journals (Sweden)

    Istihanah Nurul Eskani

    2016-04-01

    Full Text Available Kebanyakan industri batik membuang air limbah ke lingkungan tanpa diolah terlebih dahulu dengan alasan keterbatasan tempat, dana dan penguasaan teknologi. Beberapa cara pengolahan air limbah telah dilakukan untuk mengatasi penurunan mutunlingkungan akibat pembuangan air limbah.Telah dilakukan penelitian proses pengolahan air limbah batik secara kimia, biologi aerob dan biologi anaerob. Proses kimia dilaksanakan dengan menambahkan koagulan tawas dan kapur ke dalam air limbah batik. Proses biologi aerob dijalankan dalam reaktor terbuka selama 5 hari, sedang proses biologi anaerob  dijalankan dalam reaktor tertutup selama 12 hari. Hasil proses kemudian diukur parameter warna, COD dan alkalinitasnya.Hasil penelitian pengolahan air limbah batik secara kimia dapat menurunkan parameter warna yang berasal dari zat warna Naphtol sebesar 83,15 %, COD sebesar 28,82% dan pH hasil proses 7. Proses biologi anaerob menurunkan parameter warna sebesar 94,95%, COD sebesar 59,89% dan pH hasil proses 5. Proses biologi aerob dapat menurunkan parameter warna yang berasal dari zat warna Naphtol sebesar 97,82%, COD sebesar 72, 88% dan pH hasil proses 6,5. Sehingga dapat disimpulkan bahwa pengolahan limbah cair batik secara biologi aerob lebih efektif daripada pengolahan secara biologi anaerob maupaun secara kimia. Kata kunci : pengolahan air limbah batik, proses kimia, proses biologi

  4. Analisis Computational Fluid Dynamics untuk Perancangan Reaktor Gasifikasi Sekam Padi Tipe Downdraft

    Directory of Open Access Journals (Sweden)

    Dziyad Dzulfansya

    2014-10-01

    Full Text Available Rice husk is one of biomass type which can be utilized as gasification’s feedstock for producing combustible gas which can be used as fuel in internal combustion engine. The objective of this research was to obtain the best design of small scale rice husk gasifier from among geometry scenarios by applying computational fluid dynamics method. The geometry scenarios used in this study were angle of throat 70O, 80O, and 90O, and also angle of nozzel 10O and 20O. The softwares used in this study were Gambit 2.4.6 (meshing 3D model and Ansys Fluent 13.0 (simulation. The reactions involved in gasification (3 heterogeneous reactions and 6 homogeneous reactions were solved by finite rate/Eddy dissipation model. Results of simulation showed that gasifier with angle of throat 90O and angle of nozzel 10O produced the highest heating value of gas with volume fraction of CO, H2, and CH4 is 14.49%, 9.65%, and 2.39% respectively. This result showed reasonable agreement with experimental data from other researchers on rice husk gasification.

  5. A study for fuel reloading strategy in pebble bed core

    International Nuclear Information System (INIS)

    Kim, Hong Chul

    2012-02-01

    A fuel reloading analysis system for pebble bed reactor was developed by using a Monte Carlo code. The kinematic model was modified to improve the accuracy of the pebble velocity profile and to develop the model so that the diffusion coefficient is not changed by the geometry of the core. In addition, the point kernel method was employed to solve an equation derived in this study. Then, the analysis system for the pebble bed reactor was developed to accommodate the double heterogeneity, pebble velocity, and pebble refueling features using the MCNPX Monte Carlo code. The batch-tracking method was employed to simulate the movement of the pebbles and an automation system was written in the C programming language to implement it. The proposed analysis system can be utilized to verify new core analysis codes, deep-burn studies, various sensitivity studies, and other analysis tools available for the application of new fuel reloading strategies. It is noted that the proposed algorithm for the optimum fuel reloading pattern differs from other optimization methods using sensitivity analysis. In this algorithm, the reloading strategy, including the loading of fresh fuel and the reloading positions of the fresh and reloaded fuels, is determined by the interrelations of the criticality, the nuclear material inventories in the extracted fuel, and the power density. The devised algorithm was applied to the PBMR and NHDD-PBR200. The results show that the proposed algorithm can apply to satisfy the nuclear characteristics such as the criticality or power density since the pebble bed core has the characteristics that the fuels are reloaded every day

  6. The training of the staff for work with radioactive materials and work on nuclear reactor in the Institute; Obuka kadrova za rukovanje radioizotopima i pogon nuklearnih reaktora u Institutu 'Boris Kidric' - Vinca

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Mladjenovic, O; Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1978-05-15

    A short informational review of the activities in the 'Boris Kidric' Institute on the training courses for the use of radioactive materials and for operating nuclear reactors including power reactors. The survey of the courses is given in the enclosures. (author) Kratak informativni pregled delatnosti u IBK na kursevima za obuku kadrova u rukovanju readioaktivnim materijalima i pogonu nuklearnih reaktora, ukljucujuci reaktore snage. pregled kurseva i materijala za njih dati su u prilozima. (author)

  7. Inspection and repair of nuclear components

    International Nuclear Information System (INIS)

    Lahner, K.; Poetz, F.

    1993-01-01

    Despite careful design, manufacturing and operation, some of the important safety-relevant components show deterioration with time. Because of activation and contamination of these components, their inspection and repair has to be performed with manipulators. Some sophisticated manipulators are described, built by ABB Reaktor and used for inspection, maintenance and repair of PWR steam generators, fuel alignment pins, core baffle former bolts and reactor pressure vessel head penetrations. (Z.S.) 7 figs

  8. RA reactor operation and maintenance in 1990 with comparative evaluation from 1986-1990, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1990. godini, uz uporedni pregled za period 1986-1990. godina

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Vasovic, B; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1990-12-15

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The new emergency cooling system and the reconstruction of the existing ventilation system were finished in 1989, the conditions for further reactor operation were fulfilled. In the meantime new licensing regulations adopted in 1988 were not demanding the mentioned conditions for reactors operated at power less than 10MW, RA reactor power being 6.5 MW. But the reactor could not be restarted due to planned renewal of the reactor instrumentation. It is planned to exchange the complete instrumentation by the end of 1991. Training program for the staff operating and maintaining the reactor components was prepared in 1985. Reconstruction, modification and construction of components demanded new documentation needed for further safe reactor operation. New version of RA reactor safety report was finished in 1986 according to the recommendations of IAEA and licensing regulations of Yugoslavia. In 1989, new documents were written covering regulations and instructions for reactor operation. The new reactor experimental loop was designed in 1986, and constructed and tested in 1990. All the reactor components were maintained by specific reactor services. Financing of the reactor remains a permanent problem. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula 1984. godine Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Radovi na izgradnji sistema za udesno hladjenje i rekonstrukciji postojeceg sistema specijalne ventilacije zavrseni su 1989. godine. Uslovi za nastavak rada reaktora

  9. RA reactor operation and maintenance in 1989, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1989. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1989-12-15

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula 1984. Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Zavrseni su radovi na izgradnji sistema za udesno hladjenje, rekonstrukciji postojeceg sistema specijalne ventilacije i rekonstrukciji sistema za napajanje elektricnom energijom. Krajem 1988, medjunarodna agencija za atomsku energiju potpisala je ugovor sa sovjetskom firmom Atomergexport za izradu novog sistema instrumentacije. Sa ciljem da se poveca i efikasnije koristi prostor za skladistenje ozracenog goriva, 1987. godine zapoceta je realizacija projekata preciscavanja vode u bazenima za odlezavanje

  10. Annual report of the group for maintenance of electrical equipment; Godisnji izvestaj elektro grupe

    Energy Technology Data Exchange (ETDEWEB)

    Rajic, M [Odelenje odrzavanja, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-12-15

    This report includes detailed description of repairs and revisions of the electrical equipment of the RA reactor which were done according to the annual plan during the periods when reactor was not operated. Unplanned repairs are part of this report as well. [Serbo-Croat] Ovaj izvestaj sadrzi detaljan opis remontnih i revizionih radova na elektroopremi reaktora RA izvrsenih prema godisnjem planu u periodima kada reaktor nije radio. Vanplanski poslovi ukljucujuci popravke elektroopreme su takodje deo ovog izvestaja.

  11. An integrated approach to sensor FDI and signal reconstruction in HTGRs – Part I: Theoretical framework

    International Nuclear Information System (INIS)

    Uren, Kenneth R.; Schoor, George van; Rand, Carel P. du; Botha, Anrika

    2016-01-01

    Highlights: • An integrated sensor fault detection and isolation method for nuclear power plants. • Utilise techniques such as non-temporal parity space and principal component analysis. • Utilise statistical methods and fuzzy systems for sensor fault isolation. • Allow the detection of multiple sensor faults. • Proposed methodology suitable for online implementation. - Abstract: Sensor fault detection and isolation (FDI) is an important element in modern nuclear power plant (NPP) diagnostic systems. In this respect, sensor FDI of generation II and III water-cooled nuclear energy systems has become an active research topic to continually improve levels of reliability, safety, and operation. However, evolutionary advances in reactor and component technology together with different energy conversion methodologies support the investigation of alternative approaches to sensor FDI. Within this context, the basic aim of this two part series is to propose, implement and evaluate an integrated approach for sensor FDI and signal reconstruction in generation IV nuclear high temperature gas-cooled reactors (HTGRs). In part I of this two part series, the methodology and theoretical background of the integrated sensor FDI and signal reconstruction approach are given. This approach combines techniques such as non-temporal parity space analysis (PSA), principal component analysis (PCA), sensor fusion and fuzzy decision systems to form a more powerful sensor FDI methodology that exploits the strengths of the individual techniques. An illustrative example of the PCA algorithm is given making use of actual data retrieved from a pilot plant called the pebble bed micro model (PBMM). This is a prototype gas turbine power plant based on the first design configuration of the pebble bed modular reactor (PBMR). In part II, the described integrated sensor fault detection approach will be evaluated by means of two case studies. In the first case study the approach will be evaluated

  12. Cogeneration using a nuclear reactor to generate process heat

    International Nuclear Information System (INIS)

    Alonso, Gustavo; Ramirez, Ramon

    2009-01-01

    Some of the new nuclear reactor technologies (Generation III+) are claiming the production of process heat as an additional value to electricity generation. These technologies are still under development and none of them has shown how this can be possible and what will be the penalty in electricity generation to have this additional product. The current study assess the likeliness of generate process heat from a Pebble Bed Modular Reactor to be used for a refinery showing different plant balance and alternatives to produce and use that process heat. An actual practical example is presented to demonstrate the cogeneration viability using the fact that the PBMR is a modular small reactor and also the challenges that this option has. (author)

  13. Press kit. Cooperation between Areva and South Africa in the nuclear energy field

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    This document presents the nuclear industry business developed by Areva in South Africa. The first part offers general information on the country (political context, economy which fuels African growth, social situation and South Africa in search of sustainable development). An other part deals with the electricity supply (predominance of coal and the issue of global warming, electricity for everyone. The last parts detail the nuclear energy development (the new PBMR reactor project, the exploitation of all nuclear technology) and how Areva consolidates its presence in South Africa. (A.L.B.)

  14. The treatment of uranyl nitrate from the AMOR process for VKTA Rossendorf

    International Nuclear Information System (INIS)

    Boessert, W.; Krempl, R.; Miller, J.W.

    2003-01-01

    The blending of uranyl nitrate solutions at VKTA and its subsequent treatment at BNFL-Sellafield is a significant step towards the safe and effective treatment of these enriched uranyl nitrate solutions. Overall the integration of the expertise of the international company BNFL/Westinghouse will lead to the achievement of a successful solution. This success has involved the integration of the project management and operational facilities of BNFL Sellafield with the local planning, design and manufacture capacities of Westinghouse Reaktor GmbH. (orig.)

  15. Kinetics parameter measurements on RSG-GAS, a low-enriched fuel reactor

    International Nuclear Information System (INIS)

    Jujuratisbela, U; Arbie, B; Pinem, S.; Tukiran; Suparlina, L.; Singh, O.P.

    1995-01-01

    Kinetics parameter measurements, such as reactivity worths of control rods and fuel elements, beam tube void reactivity, power reactivity coefficient and xenon poisoning reactivity have been performed on different cores of Reaktor Serba Guna G.A. Siwabessy (RSG-GAS). In parallel, a programme was also initiated to measure the other kinetics parameters like effective delayed neutron life time, prompt neutron decay constant, validation of period reactivity relationship and zero power frequency response function. The paper provides the results of these measurements. (author)

  16. Nuclear Reactor RA Safety Report, Vol. 1, Introduction; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 1, Uvod

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-01

    Based on the agreement between governments of Socialist Federal Republic of Yugoslavia and USSR of January 28 1956, a contract was signed about construction of RA research reactor in the Boris Kidric Institute of nuclear sciences. Building of the RA reactor started in 1956, and has reached criticality in 1959. Since then it has been in almost permanent operation, except for five longer shutdown periods: in 1963 because of heavy water and primary coolant system contamination with cobalt; in 1970 because of transporting the heavy water to France for isotopic regeneration; in 1979/1980 and 1983 because of aluminium oxyhydrate deposition on the fuel element cladding in reactor active region; in 1985/1986 because of ventilation system reconstruction and construction of emergency core cooling system. RA reactor is a heavy water cooled and heavy water moderated research reactor. Since the beginning of its operation, 2% enriched metal uranium fuel was used. From 1976, 80% enriched uranium oxide fuel elements were used partially in some core regions and since 1981 the complete core was filled with this highly enriched fuel. RA reactor was designed to operate under normal conditions at 6.5 MW power and at 10 MW power under forced regime. As a powerful neutron source the reactor was meant to be used for research in the field of reactor and neutron physics, solid state physics, radiation chemistry, biology and radioactive isotopes production. RA reactor was build by Yugoslav companies based on USSR basic design project. Main components of the systems were produced in USSR. [Serbo-Croat] Na osnovu Sporazuma izmedju vlada SFRJ i SSSR od 28. januara 1956. godine, sklopljen je ugovor o izgradnji istrazivackog reaktora RA u Institutu za nuklearne nauke Boris Kidric u Vinci. Reaktor RA je poceo da se gradi 1956. godine, a kriticnost je postigao 1959. godine. Od tada je takoreci neprekidno bio u radu, izuzev u pet slucajeva duzeg stajanja: 1963. godine - zbog kontaminacije teske vode

  17. PENENTUAN KOEFISIEN DISPERSI ATMOSFERIK UNTUK ANALISIS KECELAKAAN REAKTOR PWR DI INDONESIA

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    by the dispersion coefficient. To find the environment dose acceptance for nuclear power plants in Indonesia, it is necessary to map the dispersion coefficient for Indonesia potential siting Model calculations in this study using Segmented plume model, which a model that is applied to the ATMOS and CONCERN module of PC-Cosyma software. The calculation has done for PWR 1000 MWe with UO2 fuel, DBA accident postulations, roughnes site conditions, for 8 example site such as Muria Peninsula, Coastal Banten, and the C, D, E, and F stability. Dispersion coefficient was calculated for the 8 fission product groups are: the noble gases, lanthanides, noble metals, halogens, alkali metals, tellurium, cerium, and strontium & barium groups. Input calculation using the program package Origen-2 and Arc View for the preparation of input calculations. The results of the dispersion parameter calculated are: the average maximum is obtained at a distance of 800 m radius from the source, for noble metals, alkali metal and cerium group nuclides. Dispersion parameters for maximum at Muria site is 1.53E-04 s/m3, Serang site is 1.40E-03 s/m3, site with stability C is 1.72E-04 s/m3, stability D is 1.40E-04 s/m3, stability E is 1.07E-04 s/m3, and site with the stability F is 2.14E-05 s/m3. Keywords: dispersion coefficient, atmospheric, PWR, accident, Indonesia

  18. Determination of bedrock depth by using microtremor array survey at the RDE Site Serpong

    International Nuclear Information System (INIS)

    Hadi Suntoko; Sriyana

    2016-01-01

    Based on the Strategic Spatial Planning (RTRW) of Tangerang Selatan 2011-2031, PUSPIPTEK Serpong has been appointed as high technology zone, therefore this area is selected as the site for Experimental Power Reactor (Reaktor Daya Eksperimental, RDE). In order to assure safety, site evaluation from several aspects have to be performed, one of which is seismological aspect. Refer to BAPETEN Chairman Regulation No. 8 year 2013, site evaluation should cover the assessment of subsurface geological condition of the site. The objective of this study is to acquire information regarding the subsurface geological condition particularly the bedrock depth of the RDE site area. Method used in this study is micro tremor array (MA). The result shows that RDE site has a bedrock type of Bojongmanik Formation at a depth of 391 meter from the surface. (author)

  19. Instruction by virtual reality to operation and security of a nuclear power plant of IV generation; Instruccion por realidad virtual a la operacion y seguridad de una central nuclear de generacion IV

    Energy Technology Data Exchange (ETDEWEB)

    Neri O, J. C.; Baltasar M, J.; Valle H, J. [Facultad de Ingenieria, Division de Estudios de Posgrado, Campus Morelos, UNAM, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico)], e-mail: neriunam@ieee.org

    2009-10-15

    The purpose of LaNuVi project which is developing in the Engineering Faculty of National Autonomous University of Mexico, to have a virtual laboratory of nuclear reactors as tool of multidisciplinary education at basic and advanced levels in nuclear engineering area, involves training resources in audio visual and interactive form that allow to form a comprehension more realistic of operation of different systems and components. In this work is proposed to use educational resources, as the employees in the U.S. Army and in some centers of advanced education of medicine, where have been come proving concepts like projected reality, increased reality, tele transparency and others that present big benefits to learning-education process. The proposal here is to include the resource knew as serious game based learning. The focal point of stage that is presented is of a nuclear reactor PBMR like desalination and generator of controlled alternating energy and efficient that should put on in operation to allow the subsistence of a community in a desolated region of beginning second quarter of X XI century. For this purpose the designs are initiated and programmed several subsystems that allow the three-dimensional modeling of main components of a PBMR as well as of surrounding facilities. The obtained results and reaches of this design are presented. The product is in tests for a first version and it is hope to achieve a free and integral resource of national distribution for different cultural groups, interested in this type of advanced technology. (Author)

  20. Instruction by virtual reality to operation and security of a nuclear power plant of IV generation

    International Nuclear Information System (INIS)

    Neri O, J. C.; Baltasar M, J.; Valle H, J.

    2009-10-01

    The purpose of LaNuVi project which is developing in the Engineering Faculty of National Autonomous University of Mexico, to have a virtual laboratory of nuclear reactors as tool of multidisciplinary education at basic and advanced levels in nuclear engineering area, involves training resources in audio visual and interactive form that allow to form a comprehension more realistic of operation of different systems and components. In this work is proposed to use educational resources, as the employees in the U.S. Army and in some centers of advanced education of medicine, where have been come proving concepts like projected reality, increased reality, tele transparency and others that present big benefits to learning-education process. The proposal here is to include the resource knew as serious game based learning. The focal point of stage that is presented is of a nuclear reactor PBMR like desalination and generator of controlled alternating energy and efficient that should put on in operation to allow the subsistence of a community in a desolated region of beginning second quarter of X XI century. For this purpose the designs are initiated and programmed several subsystems that allow the three-dimensional modeling of main components of a PBMR as well as of surrounding facilities. The obtained results and reaches of this design are presented. The product is in tests for a first version and it is hope to achieve a free and integral resource of national distribution for different cultural groups, interested in this type of advanced technology. (Author)

  1. Simulation of atmosphere stratification in the HDR test facility with the CONTAIN code

    International Nuclear Information System (INIS)

    Skerlavaj, A.; Mavko, B.; Kljenak, I.

    2001-01-01

    The test E11.2 'Hydrogen distribution in loop flow geometry', which was performed in the Heissdampf Reaktor containment test facility in Germany, was simulated with the CONTAIN computer code. The predicted pressure history and thermal stratification are in relatively good agreement with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was slightly underestimated. The analysis of simulation results enabled a better understanding of the physical phenomena during the test.(author)

  2. Radiological protection in nucleus reactor; Perlindungan radiologi di reaktor nukleus

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-31

    The chapter briefly discussed the following subjects: radiological protection problems of reactor 1. in operation 2. types of reactor i.e. power reactors, research reactors, etc. 3. during maintenance and installation of fuels. 4. nuclear fuels.

  3. Fission gas release from the sintered UO{sub 2} fuel; Oslobadjanje fisionih gasova iz goriva od sinterovanog UO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Sigulinski, F; Stevanovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1966-11-15

    This paper shoes the phenomena which control fission gases release from the sintered UO{sub 2} dependent of the burnup rate: ejection, release, diffusion, increased fission gas accumulation causing structural changes in the fuel. release of fission gases from the fuel for power reactors was studied as well. The influence of factors as temperature, characteristics of fuel, burnup rate and burnup level was analyzed. Prikazani su mehanizmi koji kontrolisu izdvajanje fisionih gasova iz sinterovanog UO{sub 2} pri razlicitim brzinama izgaranja: izletanje, izbijanje, difuzija, povecano izdvajanje fisionih gasova koje prati strukturne promene u gorivu. Razmatrano je proucavanje izdvajanja fisionih gasova iz goriva za reaktore snage. Analiziran je uticaj faktora kao sto su temperatura, karakteristike goriva, brzina i stepen izgaranja (author)

  4. Possibilities for power reactor structural material and fuel testing in reactor RA; Mogucnosti reaktora RA za testiranje konstrukcionih materijala i goriva energetskih reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Lazarevic, Dj; Stefanovic, D; Cupac, S; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-05-15

    Nuclear reactor RA at Vinca has been designed as a high flux general purpose research reactor. Among other it was intended to play a role of material testing reactor. A scope of activities of Material Laboratory and Reactor RA Department of Boris Kidric Institute is presented in this report. Reactor RA capacity for reactor structural material and fuel irradiation is also described. The increase of RA reactor irradiation capacity is based on the improvement of VISA type fuel channel for fast neutron irradiations, as well as on the general neutron flux increase, due to introduction of highly enriched uranium fuel into reactor core and the advanced in-core fuel management. The irradiation capacities described allow for the reactor material and fuel testing to the considerable extent. Istrazivacki reaktor RA u Vinci je projektovan kao visokofluksni istrazivacki reaktor opste namene. Pored ostalog, on je namenjen i za testiranje reaktorskih konstrukcionih materijala i goriva. U radu je dat pregled aktivnosti Laboratorije za materijale IBK i reaktora RA na tom podrucju, kao i opis povecanih mogucnosti reaktora RA za ozracivanje reaktorskih materijala i goriva u cilju njihovog testiranja. Povecanje mogucnosti reaktora RA zasniva se na usavrsavanju specijalnog gorivnog kanala tipa VISA (za ozracivanje materijala brzim neutronima), kao i na opstem povecanju neutronskog fluksa na osnovu uvodjenja i nacina koriscenja visokoobogacenog uranskog goriva u reaktoru RA. Opisane mogucnosti reaktora RA dozvoljavaju u znatnoj meri ispitivanje konstrukcionih materijala i goriva energetskih reaktora.

  5. Atomic Energy Research benchmark activity

    International Nuclear Information System (INIS)

    Makai, M.

    1998-01-01

    The test problems utilized in the validation and verification process of computer programs in Atomic Energie Research are collected into one bunch. This is the first step towards issuing a volume in which tests for VVER are collected, along with reference solutions and a number of solutions. The benchmarks do not include the ZR-6 experiments because they have been published along with a number of comparisons in the Final reports of TIC. The present collection focuses on operational and mathematical benchmarks which cover almost the entire range of reaktor calculation. (Author)

  6. Towards rational design of redox-stratified biofilms

    DEFF Research Database (Denmark)

    Lackner, Susanne

    Biologisk kvælstoffjernelse er en central proces indenfor avanceret spildevandsrensning. Denne afhandling beskriver anvendelsen af beluftede membran biofilm reaktorer (eng: MABRs) til fuld autotrof kvælstoffjernelse. Denne ret nye kvælstofomsætningsvej baseres på delvis omdannelse af ammonium til....... Som illustreret ved modelundersøgelserne, kan biofilm afrivning påvirke rektorfunktionen alvorligt. Den udførte forsøg for at vurdere om kemisk modificering af biofilmoverfladen kan øge dens modstandsdygtighed overfor afrivning og give forbedret biofilm tykkelseskontrol. Alt i alt kan MABR, hvor ilt...

  7. Integration of nuclear energy into oil sands projects - HTR2008-58239

    International Nuclear Information System (INIS)

    Finan, A.; Kadak, A. C.

    2008-01-01

    support further study. A single 500 MWth PBMR reactor is able to supply high-pressure steam for a 40, 000 to 60, 000 bpd Steam Assisted Gravity Drainage (SAGD) plant, whereas the CANDU and ACR reactors are unable to produce sufficient steam pressures to be practical in that application. The CANDU, ACR and PBMR reactors have potential for supplying heat and electricity for surface mining operations. The primary environmental benefit of nuclear energy in this application is to reduce CO 2 emissions by up to 3.1 million metric tons per year for each 100, 000 barrel per day (bpd) bitumen production SAGD facility, or 2.0 million metric tons per year for the replacement of 700 MWe of grid electricity with a nuclear power plant. Should carbon emissions be priced, the economic advantages of nuclear energy would be dramatically improved such that with a $50/ton CO 2 e at the releases expected for typical projects using natural gas, breakeven gas prices for nuclear drop to less than $3.50/MMBtu, well below the current natural gas price of $10/MMBtu for SAGD steam production. (authors)

  8. Turbulence-induced heat transfer in PBMR core using LES and RANS

    International Nuclear Information System (INIS)

    Lee, Jung-Jae; Yoon, Su-Jong; Park, Goon-Cherl; Lee, Won-Jae

    2007-01-01

    This paper introduces the results of numerical simulations on flow fields and relevant heat transfer in the pebble bed reactor (PBR) core, since the coolant passes a highly complicated random flow path with a high Reynolds number, an appropriate treatment of the turbulence is required. A set of simple experiments for the flow over a circular cylinder with heat transfer was conducted to finally select the large eddy simulation (LES) and k-ω model among the considering Reynolds-averaged Navier-Stokes (RANS) models for PBR application. Using these models, the PBR cores, whose geometries were simplified to the body-centered cubical (BCC) and face-centered cubical (FCC) structures, were simulated. A larger pressure drop, a more random flow field, a higher vorticity magnitude and a higher temperature at the local hot spots on the pebble surface were found in the results of the LES than in those of RANS for both geometries. In cases of the LES, the flow structures were resolved up to the grid scales. Irregular distributions of the flow and local heat transfer were found in the BCC core, while relatively regular distributions for the FCC core. The turbulent nature of the coolant flow in the pebble core evidently affected the fuel surface temperature distribution. (author)

  9. Radioactive waste management plan for the PBMR (Pty) Ltd fuel plant

    International Nuclear Information System (INIS)

    Makgae, Mosidi E.

    2009-01-01

    The Pebble Bed Modular Reactor (Pty) Ltd Fuel Plant (PFP) radioactive waste management plan caters for waste from generation, processing through storage and possible disposal. Generally, the amount of waste that will be generated from the PFP is Low and Intermediate Level Waste. The waste management plan outlines all waste streams and the management options for each stream. It also discusses how the Plant has been designed to ensure radioactive waste minimisation through recycling, recovery, reuse, treatment before considering disposal. Compliance to the proposed plan will ensure compliance with national legislative requirements and international good practice. The national and the overall waste management objective is to ensure that all PFP wastes are managed appropriately by utilising processes that minimize, reduce, recover and recycle without exposing employees, the public and the environment to unacceptable impacts. Both International Atomic Energy Agency (IAEA) and Department of Minerals and Energy (DME) principles act as a guide in the development of the strategy in order to ensure international best practice, legal compliance and ensuring that the impact of waste on employees, environment and the public is as low as reasonably achievable. The radioactive waste classification system stipulated in the Radioactive Waste Management Policy and Strategy 2005 will play an important role in classifying radioactive waste and ensuring that effective management is implemented for all waste streams, for example gaseous, liquid or solid wastes.

  10. PENURUNAN COD, TSS DAN TOTAL FOSFAT PADA SEPTIC TANK LIMBAH MATARAM CITRA SEMBADA CATERING DENGAN MENGGUNAKAN WASTEWATER GARDEN (Degradation of COD, TSS and Total Phosphate in Septic Tank Wastewater of Mataram Citra Sembada Catering Using Wastewater

    Directory of Open Access Journals (Sweden)

    Dradjat Suhardjo

    2008-07-01

    Full Text Available ABSTRAK  Sumber limbah berasal dari septictank industri restauran (catering Citra Sembada Catering, termasuk dalam kategori limbah domestik. Limbah tersebut banyak mengandung komponen yang tidak diinginkan bila dibuang ke badan air. Konsentrasi limbah yang masih di atas baku mutu, di antaranya akan memunculkan masalah pencemaran. Reaktor Wastewater Garden yang menggunakan krikil (0,5Cm-1cm dan 6 jenis tanaman yaitu : melati air (Echinodoras paleafias, Cyperus (Cyperus, Futoi (Hippochaetes lymnenalis, Pisang air (Typhonodorum indleyanum, Pickerel rush (Pontedoria cordata, Cattail (Typha latifulia. Penelitian ini bertujuan untuk mengetahui tingkat efektivitas reaktor Wastewater Garden, apabila digunakan untuk menurunkan konsentrasi Chemical Oxygen Demand (COD, Total Suspended Solid (TSS dan Fosfat Total sebagai faktor pencemar pada limbah industri restauran (Citra Sembada Catering yang tertampung pada septictank. Penelitian dilakukan dengan menggunakan reaktor Wastewater Garden dengan sistem batch dan dimensi reaktor lm x 0.5m x lm. Zona air limbah 75 cm, dan zona substrat atau krikil 80 cm, akar tanaman ditanam sedalam l0-15 cm. Metode penelitian yang digunakan berdasarkan SNI, di mana COD mengacu pada SNI 06-6989.2-2004 metode refluks tertutup secara spektrofotometri, TSS mengacu pada SK SNI M-03-1990-F metode pengujian secara gravimetri dan Fosfat total mengacu pada SNI M-52-1990-03 metode asam askorbat dengan alat spektrofotometer. Penelitian ini dilakukan selama 12 hari di mana setiap 3 hari sampel diambil pada outlel kemudian dianalisis. Berdasarkan hasil penelitian ini, diperoleh bahwa penggunaan wastewater garden pada limbah cair Mataram Citra Sembada Catering dapat menurunkan COD dengan efektivitas optimum 40,81% pada hari ke-6, penurunan TSS 89,l2% pada efektifitas optirnum hari ke-12 dan penurunan fosfat total dengan efektivitas optimum pada hari ke-6 yaitu sebesar 99,73 %. Tanaman dapat hidup dengan subur.   ABSTRACT  Wastewater

  11. Operation and maintenance of the RA reactor, RA Research reactor. Annual report 1976; Pogon i odrzavanje reaktora RA, Izvestaj o radu u 1976. godini

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    During 1976 the Ra reactor was operating for about 30% shorter period than usual. The reason were extraordinary repair activities within regular and investment maintenance as well as repair of failures caused by neglected maintenance during previous 6 years. Delay was caused by unavailability of fuel (2% enriched fuel elements are spent) and the new 80% enriched fuel demanded experimental and theoretical analyses before being introduced into the core. Safety analyses concerned with using 80% enriched fuel both experimental and theoretical were successfully fulfilled. The December 1976 successful experimental campaign can be marked as end of the 17 years period of using 2% enriched fuel and start of the new period of using highly enriched fuel. This is significant not only for the reactor itself but for the users, because it would result in increase of neutron flux by 50% with the increase of costs by only 4%. Demand was submitted for obtaining the final license for transition operating regime with highly enriched fuel which would save at least 2 200 000 dinars. This will enable reactor operation in 1977 and later on, without interruption by 'critical' and other experiments related to new highly enriched fuel. A high number of repair and other urgent activities were fulfilled in order to enable safe operation. Some of these activities were done never before and some were neglected during past 6 years. The most important tasks were: purchase of Al tubes made of special alloy, fabrication and mounting of the fuel channel; overall investigation of reactor vessel leakage; repair of the heavy water pump; exchange of two vertical channels. basic equipment for construction of emergency cooling system was purchased. Hot cells are equipped for independent utilisation. [Serbo-Croat] Reaktor RA je u 1976. godini radio za oko 30% krace vreme od uobicajenog. Razlog su bill izuzetno veliki obim remontnih i drugih radova u okviru tekuceg i investicionog odrzavanja, kao i krupnih

  12. PENGURANGAN KADAR CO2 MENGGUNAKAN SPIRULINA PLATENSIS DALAM TUBULAR BIOREACTOR

    Directory of Open Access Journals (Sweden)

    Zainal Syam Arifin

    2015-06-01

    Full Text Available Increasing the population impact on increasing energy demand. On the other hand, the energy generation industry has been blamed as one of the contributors of carbon dioxide about 25% of total CO2 emissions worldwide. Meanwhile, the production of biogas, which aims to address the increasing need of energy, produces carbon dioxide in the range of 25–50% by volume. To overcome this, a cheap method, optimum and efficient as well as environmentally friendly in reducing CO2 levels by using Spirulina platensis is needed. This research aims to created a mathematical models and found the optimum flow rate to reduced levels of CO2 by using Spirulina platensis. This study used a glass tubular bioreactor (D = 2.6 cm at a temperature of 30°C and irradiated with a fluorescent lamp Philips TL 36 Watt, color temperature: 6,200K cool daylight, light output: 2,600 lm, 72 lm/W. Tubular reactor was placed in a box lined with silver foil walls on three sides. With mathematical models of tubular reactor, the reaction rate constants could be predicted. Based on calculations of data and graphs, optimum volumetric velocity could also be predicted. Variation of flowrate to observed the reduction rate of CO2 was 0.25 mL/sec, 0.35 mL/sec, 0.5 mL/sec, 0.75 mL/sec, 1 mL/sec. Carbon source was 99.99% CO2.Observations of Spirulina growth was made on the flow rate of 0.25 mL/sec at the initial levels of dry weight 2.1208 g/L. The results of this study indicated that the low flowratewas a more effective way to reduced carbon dioxide levels using Spirulina platensis (= 2.82×10-4 sec-1. The highest conversion was obtained at a volumetric flow rate of 0.25 mL/sec and optimum speeds in the range of 0.3 to 0.4 mL/sec. The rate of incoming CO2 flux should be less than 0.047 mL/cm2.detik. Specific Growth Rate (µ of Spirulina platensis in this study was 2.56×10-2 minute-1.   Keywords: Spirulina platensis, a vertical tubular bioreactor, CO2 reduction     ABSTRAK

  13. Power conversion unit for the South African direct cycle HTGR

    International Nuclear Information System (INIS)

    Liebenberg, J.J.

    1997-01-01

    The system parameters chosen to optimise the thermal efficiency of the Eskom PBMR whilst maintaining component simplicity is discussed. Power Conversion Unit components, which are now at a preliminary design stage comprise a precooler, two turbo units consisting of a turbine driven compressor, recuperator and a power turbine, driving an alternator. Design aspects of every component is mentioned and the inventory method of poorer control is explained with reference to start-up and and shut-down events, the system an effective load following device, down to 4% of full power. Application of the same design principles for HTGRs smaller than 25 MWe is discussed. (author)

  14. Development of technologies for nuclear reactors of small and medium sized

    International Nuclear Information System (INIS)

    2011-08-01

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  15. Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-08-15

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  16. Development of computational methods for the safety assessment of gas-cooled high-temperature and supercritical light-water reactors. Final report; Rechenmethoden zur Bewertung der Sicherheit von gasgekuehlten Hochtemperaturreaktoren und superkritischen Leichtwasserreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, S.; Cron, D. von der; Hristov, H.; Lerchl, G.; Papukchiev, A.; Seubert, A.; Sureda, A.; Weis, J.; Weyermann, F.

    2012-12-15

    This report documents developments and results in the frame of the project RS1191 ''Development of computational methods for the safety assessment of gas-cooled high temperature and supercritical light-water reactors''. The report is structured according to the five work packages: 1. Reactor physics modeling of gas-cooled high temperature reactors; 2. Coupling of reactor physics and 3-D thermal hydraulics for the core barrel; 3. Extension of ATHLET models for application to supercritical reactors (HPLWR); 4. Further development of ATHLET for application to HTR; 5. Further development and validation of ANSYS CFX for application to alternative reactor concepts. Chapter 4 describes the extensions made in TORT-TD related to the simulation of pebble-bed HTR, e.g. spectral zone buckling, Iodine-Xenon dynamics, nuclear decay heat calculation and extension of the cross section interpolation algorithms to higher dimensions. For fast running scoping calculations, a time-dependent 3-D diffusion solver has been implemented in TORT-TD. For the PBMR-268 and PBMR-400 as well as for the HTR-10 reactor, appropriate TORT-TD models have been developed. Few-group nuclear cross sections have been generated using the spectral codes MICROX- 2 and DRAGON4. For verification and validation of nuclear cross sections and deterministic reactor models, MCNP models of reactor core and control rod of the HTR-10 have been developed. Comparisons with experimental data have been performed for the HTR-10 first criticality and control rod worth. The development of the coupled 3-D neutron kinetics and thermal hydraulics code system TORT-TD/ATTICA3D is documented in chapter 5. Similar to the couplings with ATHLET and COBRA-TF, the ''internal'' coupling approach has been implemented. Regarding the review of experiments and benchmarks relevant to HTR for validation of the coupled code system, the PBMR-400 benchmarks and the HTR-10 test reactor have been selected

  17. Current status and future development of modular high temperature gas cooled reactor technology

    International Nuclear Information System (INIS)

    2001-02-01

    This report includes an examination of the international activities with regard to the development of the modular HTGR coupled to a gas turbine. The most significant of these gas turbine programmes include the pebble bed modular reactor (PBMR) being designed by ESKOM of South Africa and British Nuclear Fuels plc. (BNFL) of the United Kingdom, and the gas turbine-modular helium reactor (GT-MHR) by a consortium of General Atomics of the United States of America, MINATOM of the Russian Federation, Framatome of France and Fuji Electric of Japan. Details of the design, economics and plans for these plants are provided in Chapters 3 and 4, respectively. Test reactors to evaluate the safety and general performance of the HTGR and to support research and development activities including electricity generation via the gas turbine and validation of high temperature process heat applications are being commissioned in Japan and China. Construction of the high temperature engineering test reactor (HTTR) by the Japan Atomic Energy Research Institute (JAERI) at its Oarai Research Establishment has been completed with the plant currently in the low power physics testing phase of commissioning. Construction of the high temperature reactor (HTR-10) by the Institute of Nuclear Energy Technology (INET) in Beijing, China, is nearly complete with initial criticality expected in 2000. Chapter 5 provides a discussion of purpose, status and testing programmes for these two plants. In addition to the activities related to the above mentioned plants, Member States of the IWGGCR continue to support research associated with HTGR safety and performance as well as development of alternative designs for commercial applications. These activities are being addressed by national energy institutes and, in some projects, private industry, within China, France, Germany, Indonesia, Japan, the Netherlands, the Russian Federation, South Africa, United Kingdom and the USA. Chapter 6 includes details

  18. Action plan during reactor shutdown in October 1965, Annex 5; Prilog br. 5 - Plan radova u toku stajanja reaktora u mesecu oktobru 1965. godine

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M [Reaktor RA, Odelenje odrzavanja, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1965-12-15

    The action plan of the division for reactor maintenance during reactor shutdown includes detailed list of tasks for mechanics, electronic and electrical equipment group during the reactor shutdown period in October 1965. It contains tasks for planned shutdown periods in September, August, July, May, April, March, and February 1965. [Serbo-Croat] Plan radova Odelenja odrzavanja reaktora RA za period stajanja reaktora u oktobru mesecu 1965. sadrzi detaljnu listu zadataka masinske grupe, elektro grupe i elektronske grupe. Ovaj prilog sadrzi i zadatke koji ce biti obavljeni tokom planiranih perioda kada je reaktor zaustavljen u septembru, avgustu, julu, junu, maju, aprilu, martu i februaru 1965.

  19. PENGOLAHAN LIMBAH CAIR ORGANIK SECARA BIOLOGI MENGGUNAKAN REAKTOR ANAEROBIK LEKAT DIAM

    OpenAIRE

    Indriyati, Indriyati

    2018-01-01

       Organic waste water can be treated biolocally by using anaerobic fixed bed reactor. Fixed bed reactor is bioreactor which is compleeted with support material inside reactor for bacteria fixation in the surface area of support material. The benefit of using this kind of technology are it needs low energy, low nutrien, low sludge production and could treat high organic concentraion waste water.   The support material  has important role in the  Fixed Bed reactor performance, therefore it mus...

  20. Economic evaluation of nuclear seawater desalination in an Algerian site: La Macta

    International Nuclear Information System (INIS)

    Belkaid, Abderrahmane; Amzert, Sid Ahmed; Arbaoui, Fahd; Bouaichaoui, Youcef

    2010-01-01

    As the needs for fresh water and electricity increases rapidly in Algeria, the Algerian authorities launched a study to assess the potentialities of the introduction of nuclear energy for the production of electricity and potable water. This study which started in 2007, is held under the framework of an IAEA Project untitled: 'techno-economic feasibility study of seawater desalination using nuclear energy' and its objective is to provide a document which will be used to support the government's decision to introduce the nuclear desalination in Algeria. To that end, one site has been selected to host nuclear desalination plant. This site is located in North West region of the country. In this study, we present the results achieved under this project and which corresponds to the economical evaluation of coupling several nuclear reactors: GT-MHR, PBMR, AP1000 and PWR900, with two desalination processes MED and RO. The results are compared with those obtained with fossil energy sources: Natural Gas Turbine and Natural Gas Combined Cycle. (author)

  1. Effects of phosphorous-doping and high temperature annealing on CVD grown 3C-SiC

    International Nuclear Information System (INIS)

    Rooyen, I.J. van; Neethling, J.H.; Henry, A.; Janzén, E.; Mokoduwe, S.M.; Janse van Vuuren, A.; Olivier, E.

    2012-01-01

    The integrity and property behavior of the SiC layer of the Tri-isotropic (TRISO) coated particle (CP) for high temperature reactors (HTR) are very important as the SiC layer is the main barrier for gaseous and metallic fission product release. This study describes the work done on un-irradiated SiC samples prepared with varying phosphorus levels to simulate the presence of phosphorus due to transmutation. 30 Si transmutes to phosphorous ( 31 P) and other transmutation products during irradiation, which may affect the integrity of the SiC layer. The P-doping levels of the SiC samples used in this study cover the range from 1.1 × 10 15 to 1.2 × 10 19 atom/cm 3 and are therefore relevant to the PBMR operating conditions. Annealing from 1000 °C to 2100 °C was performed to study the possible changes in nanostructures and various properties due to temperature. Characterization results by X-ray diffraction (XRD), secondary ion mass spectrometry (SIMS), scanning electron microscopy (SEM), transmission electron microscopy (TEM) and high resolution transmission electron microscopy (HRTEM), are reported in this article. As grain boundary diffusion is identified as a possible mechanism by which 110m Ag, one of the fission activation products, might be released through intact SiC layer, grain size measurements is also included in this study. Temperature is evidently one of the factors/parameters amongst others known to influence the grain size of SiC and therefore it is important to investigate the effect of high temperature annealing on the SiC grain size. The ASTM E112 method as well as electron back scatter diffraction (EBSD) was used to determine the grain size of various commercial SiC samples and the SiC layer in experimental PBMR Coated Particles (CPs) after annealing at temperatures ranging from 1600 °C to 2100 °C. The HRTEM micrograph of the decomposition of SiC at 2100 °C are shown and discussed. Nanotubes were not identified during the TEM and HRTEM analysis

  2. Effects of phosphorous-doping and high temperature annealing on CVD grown 3C-SiC

    Energy Technology Data Exchange (ETDEWEB)

    Rooyen, I.J. van, E-mail: Isabella.vanrooyen@inl.gov [CSIR, National Laser Centre, PO Box 395, Pretoria 0001 (South Africa); Department of Physics, Nelson Mandela Metropolitan University, PO Box 77000, Port Elizabeth 6031 (South Africa); Fuel Design, PBMR, 1279 Mike Crawford Avenue, Centurion 0046 (South Africa); Neethling, J.H. [Department of Physics, Nelson Mandela Metropolitan University, PO Box 77000, Port Elizabeth 6031 (South Africa); Henry, A.; Janzen, E. [Department of Physics, Chemistry and Biology, Semiconductor Materials, Linkoeping University, Linkoeping 58183 (Sweden); Mokoduwe, S.M. [Fuel Design, PBMR, 1279 Mike Crawford Avenue, Centurion 0046 (South Africa); Janse van Vuuren, A.; Olivier, E. [Department of Physics, Nelson Mandela Metropolitan University, PO Box 77000, Port Elizabeth 6031 (South Africa)

    2012-10-15

    The integrity and property behavior of the SiC layer of the Tri-isotropic (TRISO) coated particle (CP) for high temperature reactors (HTR) are very important as the SiC layer is the main barrier for gaseous and metallic fission product release. This study describes the work done on un-irradiated SiC samples prepared with varying phosphorus levels to simulate the presence of phosphorus due to transmutation. {sup 30}Si transmutes to phosphorous ({sup 31}P) and other transmutation products during irradiation, which may affect the integrity of the SiC layer. The P-doping levels of the SiC samples used in this study cover the range from 1.1 Multiplication-Sign 10{sup 15} to 1.2 Multiplication-Sign 10{sup 19} atom/cm{sup 3} and are therefore relevant to the PBMR operating conditions. Annealing from 1000 Degree-Sign C to 2100 Degree-Sign C was performed to study the possible changes in nanostructures and various properties due to temperature. Characterization results by X-ray diffraction (XRD), secondary ion mass spectrometry (SIMS), scanning electron microscopy (SEM), transmission electron microscopy (TEM) and high resolution transmission electron microscopy (HRTEM), are reported in this article. As grain boundary diffusion is identified as a possible mechanism by which {sup 110m}Ag, one of the fission activation products, might be released through intact SiC layer, grain size measurements is also included in this study. Temperature is evidently one of the factors/parameters amongst others known to influence the grain size of SiC and therefore it is important to investigate the effect of high temperature annealing on the SiC grain size. The ASTM E112 method as well as electron back scatter diffraction (EBSD) was used to determine the grain size of various commercial SiC samples and the SiC layer in experimental PBMR Coated Particles (CPs) after annealing at temperatures ranging from 1600 Degree-Sign C to 2100 Degree-Sign C. The HRTEM micrograph of the decomposition of Si

  3. New advanced small and medium nuclear power reactors: possible nuclear power plants for Australia

    International Nuclear Information System (INIS)

    Dussol, R.J.

    2003-01-01

    In recent years interest has increased in small and medium sized nuclear power reactors for generating electricity and process heat. This interest has been driven by a desire to reduce capital costs, construction times and interest during construction, service remote sites and ease integration into small grids. The IAEA has recommended that the term 'small' be applied to reactors with a net electrical output less than 300 MWe and the term 'medium' to 300-700 MWe. A large amount of experience has been gained over 50 years in the design, construction and operation of small and medium nuclear power reactors. Historically, 100% of commercial reactors were in these categories in 1951-1960, reducing to 21% in 1991-2000. The technologies involved include pressurised water reactors, boiling water reactors, high temperature gas-cooled reactors, liquid metal reactors and molten salt reactors. Details will be provided of two of the most promising new designs, the South African Pebble Bed Modular Reactor (PBMR) of about 110 MWe, and the IRIS (International Reactor Innovative and Secure) reactor of about 335 MWe. Their construction costs are estimated to be about US$l,000/kWe with a generating cost for the PBMR of about US1.6c/kWh. These costs are lower than estimated for the latest designs of large reactors such as the European Pressurised Reactor (EPR) designed for 1,600 MWe for use in Europe in the next decade. It is concluded that a small or medium nuclear power reactor system built in modules to follow an increasing demand could be attractive for generating low cost electricity in many Australian states and reduce problems arising from air pollution and greenhouse gas emissions from burning fossil fuels

  4. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor; Istrazivacki nuklearni reaktor RA - Deo 1 - Pogon, odrzavanje i eksploatacija nuklearnog reaktora RA u 1982. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R; Kozomara-Maic, S; Cupac, S; Radivojevic, J; Stamenkovic, D; Skoric, M; Miokovic, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1982-12-15

    outdated spare parts. Project concerning renewal of the reactor equipment was initiated during the past year according to the contract with the Soviet Atomenergoexport and IAEA which has planned to spend 1 000 000 of rubles for this project. [Serbo-Croat] Probni rad reaktora zapocet septembra 1981. godine na snazi od 2 MW sa 80% obogacenim gorivom nastavljen je u celoj 1982. prema prethodno napravljenom planu. Pocetno jezgro formirano ja sa 44 gorivna kanala sa po deset gorivnih elemenata. Prva polovina godine iskoriscena je za neophodna merenja i ispitivanja radnih parametara rektora i funkcionisanje sistema i opreme u radnim uslovima. U drugoj polovini godine zapocet je program probnog rada na visim snagama. Utvrdjeno je da ugradjeni visak reaktivnosti i kapacitet kontrolnih sipki zadovoljava sigurnosne kriterijume MAAE, ugradjeni visak reaktivnosti moze da omoguci rad na snazi od 4,7 MW u 4 mesecne kampanje sa po 15-20 dana rada, postoje povoljni uslovi za hladjenje jezgra pri pocetnoj konfiguraciji. Izmeren je efekat pocetnog zatrovanja na reaktivnost i raspodelu snage, izmerena je pocetna prostorna raspodela neutronskog fluksa koja iznosi 3,9 10{sup 13} cm{sup -2} s{sup -1} pri znazi od 2 MW. Odredjena je promena kalibracionog koeficijenta u sistemu za automatsko odrzavanje snage. Svi rezultati ukazuju da ce pri nominalnoj snazi od 4,7 Mw biti zadovoljeni svi kriterijum sigurnosti i postovana ogranicenja u odnosu na koriscenje goriva. Po dobijanju dozvole za rad na punoj snazi morace da se izvrsi dopunski probni rad na snagama od 3, 4, i 4,7 MW. Prelaz od pocetne konfiguracije sa 44 gorivna kanala u jezgru vrsice se postupno da bi se dostigla ravnotezna konfiguracija sa 72 gorivna kanala sa po 10 elemenata. Reaktor nije radio u septembu mesecu zbog radova na zameni dela cevovoda koji povezuje pumpnu stanicu na Dunavu sa horizontalnim taloznikom. Kontrola i odrzavanje opreme izvrsavani su redovno u granicama raspolozivosti rezerbinh delova. Teskocu pricinjava

  5. UNJUK KERJA REAKTOR PLASMA DIELECTRIC BARRIER DISCHARGE UNTUK PRODUKSI BIODIESEL DARI MINYAK KELAPA SAWIT

    Directory of Open Access Journals (Sweden)

    Ardian Dwi Yudhistira

    2013-10-01

    Full Text Available Biodiesel is one of alternative renewable energy source to substitute diesel fuel. Various biodiesel productionprocesses through transesterification reaction with a variety of catalysts have been developed by previousresearcher. This process still has the disadvantage of a long reaction time, and high energy need. DielectricBarrier Discharge (DBD plasma electro-catalysis may become a solution to overcome the drawbacks in theconventional transesterification process. This process only needs a short time reaction and low energy process.The purpose of this study was to assess the performance of DBD plasma rector in making biodiesel such as: theeffect of high voltage electric value, electrodes gap, mole ratio of methanol / oil, and reaction time. TheResearch method was using GC-MS (Gas Cromatography-Mass Spectrofotometry and FTIR (FourierTransform Infrared Spectrofotometry and then it will be analysed the change of chemical bond betweenreactant and product. So, the reaction mechanism can be predicted. Biodiesel is produced using methanol andpalm oil as reactants and DBD plasma used as reactor in batch system. Then, reactants contacted by highvoltage electric. From the results of this research can be concluded that the reaction mechanism occurs in theprocess is the reaction mechanism of cracking, the higher of electric voltage and the longer of reaction time leadto increasing of product yield. The more of mole ratio of methanol / oil and widening the gap between theelectrodes lead to decreased product yield. From this research, product yield maksimum is 89,8% in the variableof rasio mol metanol/palm oil 3:1, voltage 10 kV, electrode gap 1,5 cm, and reaction time 30 seconds.

  6. The development of quality assurance program in Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Rosli Darmawan; Mohd Rizal Mamat; Mohamad Zaid Mohamad; Mohd Ridzuan Abdul Mutalib

    2007-01-01

    One of the trivial issues in the operation of Nuclear Reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing has always bring about general public fear on anything related to nuclear. IAEA has always emphasized on the assurance of nuclear safety for all nuclear installations and activities. According to the IAEA safety guides, all research reactors are required to implement quality assurance programs to ensure the conduct of operations are in accordance with the safety standards required. This paper discusses the activities carried out toward the establishment of Quality Assurance Program for Reaktor TRIGA PUSPATI (RTP). (Author)

  7. Design Of Pump Monitoring Of Primary Cooling System

    International Nuclear Information System (INIS)

    Indrakoesoema, Koes; Sujarwono

    2000-01-01

    Monitoring of 3 primary cooling pumps done visually by operator on the spot. The operator must be check oil in a sight glass, oil leakage during pump operation and water leakage. If reaktor power increase about more than 3 MW, the radiation exposure also increase in the primary cell and that's way the operator can not check the pumps. To continuing monitor all pump without delay, one system has been added I.e Closed Circuit Television (CCTV). This system using 3 video camera to monitor 3 pumps and connected to one receiver video monitor by coaxial cable located in Main Control Room. The sequence monitoring can be done by sequential switcher

  8. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H.

    2005-01-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600

  9. HTR-E project. High-temperature components and systems

    International Nuclear Information System (INIS)

    Breuil, E.; Exner, R.

    2002-01-01

    The HTR-E European project (four years project) is proposed for the 5th Framework Programme and concerns the technical developments needed for the innovative components of a modern HTR with a direct cycle. These components have been selected with reference to the present projects (GT-MHR, PBMR): (1) the helium turbine, the recuperator heat exchanger, the electro-magnetic bearings and the helium rotating seal; (2) the tribology. Sliding innovative components in helium environment are particularly concerned. (3) the helium purification system. Recommendations on impurities contents have to be provided in accordance with the materials proposed for the innovative components. The main outcomes expected from the HTR-E project are the design recommendations and identification of further R and D needs for these components. This will be based: (1) on experience feedback from European past helium test loops and reactors; (2) on design studies, thermal-hydraulic and structural analyses; (3) and on experimental tests

  10. German Light-Water-Reactor Safety-Research Program

    International Nuclear Information System (INIS)

    Seipel, H.G.; Lummerzheim, D.; Rittig, D.

    1977-01-01

    The Light-Water-Reactor Safety-Research Program, which is part of the energy program of the Federal Republic of Germany, is presented in this article. The program, for which the Federal Minister of Research and Technology of the Federal Republic of Germany is responsible, is subdivided into the following four main problem areas, which in turn are subdivided into projects: (1) improvement of the operational safety and reliability of systems and components (projects: quality assurance, component safety); (2) analysis of the consequences of accidents (projects: emergency core cooling, containment, external impacts, pressure-vessel failure, core meltdown); (3) analysis of radiation exposure during operation, accident, and decommissioning (project: fission-product transport and radiation exposure); and (4) analysis of the risk created by the operation of nuclear power plants (project: risk and reliability). Various problems, which are included in the above-mentioned projects, are concurrently studied within the Heiss-Dampf Reaktor experiments

  11. Start-up and Performance Characteristics of a Trickle Bed Reaktor Degrading Toluene

    Czech Academy of Sciences Publication Activity Database

    Misiaczek, O.; Paca, J.; Halecký, M.; Gerrard, A. M.; Sobotka, Miroslav; Soccol, C. R.

    2007-01-01

    Roč. 50, č. 5 (2007), s. 871-877 ISSN 1516-8913 Grant - others:GA ČR(CZ) GA104/05/0194 Institutional research plan: CEZ:AV0Z50200510 Source of funding: V - iné verejné zdroje Keywords : biotrickling filter * constructed mixed population * toluene Subject RIV: EE - Microbiology, Virology Impact factor: 0.349, year: 2007

  12. The recent activities of nuclear power globalization. Our provision against global warming by global deployment of our own technologies as integrated nuclear power plant supply company'

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi; Suzuki, Shigemitsu

    2008-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) is striving to expand and spread nuclear power plants as an 'Integrated Nuclear Power Plant Supply Company' based on its engineering, manufacturing, and technological support capabilities. The company also has ample experience in the export of major components. MHI is accelerating its global deployment through the market introduction of large-sized strategic reactor US-APWR, the joint development of a mid-sized strategic reactor ATMEA1 with AREVA, and a small strategic reactor PBMR. The company also plans to internationally deploy technologies for the nuclear fuel cycle. We present here the leading-edge trends in the global deployment of these nuclear businesses, all of which help to solve the energy and environmental issues in the world. (author)

  13. RA reactor operation and maintenance, Annual report 1974; Pogon i odrzavanje reaktora RA - Izvestaj o radu u 1974. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, D et al [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1974-12-15

    During 1974, RA reactor was operated at nominal power for 194 days and 13 days at lower power levels. The total production was 30711 MWh which is 2.4% higher than planned. Practically there were no discrepancies from the plan. The reactor was used for irradiation and experiments according to the demand of 437 users. This report contains detailed data about reactor power and experiments performed in 1974. Total number of afety shutdowns was 11, of which 8 were caused by power cuts, and 3 due to human error. Maximum individual personnel exposure dose was 50% of the maximum permissible dose. There were no accidents during this year. Decontamination of surfaces was less than during previous years. About 805 m{sup 2} of surfaces and 178 objects were decontaminated. It was concluded that the successful operation in 1974 has a special significance taking into account the financial problems. [Serbo-Croat] Reaktor RA je u 1974. godini radio na nominalnoj snazi 194 dana i 13 dana na manjim snagama. Ukupni rad iznosio je 30711 MWh odnosno 1,4% vise od planiranog. Prakticno nije bilo odstupanja od plana rada. Reaktor je koriscen za ozracivanja i eksperimente za 437 korisnika. Ovaj izvestaj sadrzi detaljne podatke o radu i eksperimentima koji su obavljani. U toku godine bilo je 11 sigurnosnih zaustavljanja, od cega 8 zbog elektricnog napona i 3 usled greske osoblja. Ukupna doza ozracivanja ljudstva bila je manja nego prethodnih godina. Maksimalna doza po oveku bila je 50% manja od maksimalno dozvoljene doze. Nije bilo ni jednog akcidenta. Dekontaminirano je znatno manje povrsina nego ranijih godina, i sakupljeno manje otpada nego prethodnih godina, dok tecnih efluenata nije bilo. Zakljuceno je da uspesan rad reaktora u 1974. godini ima poseban znacaj kada se imaju na umu problemi finansiranja reaktora.

  14. Report on the activity of the RA reactor operation for the period from July 1 1961 - Sept. 30 1961; Tromesecni izvestaj za period od 1.VII do 30.IX 1961. g

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Laboratorija za eksploataciju reaktora RA, Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-09-15

    During the reporting period the reactor was permanently ready for operation and responding to the demands of the experimenters. The reactor was operating for 408.5 hours at power levels from 50 - 5000 kW, or 1985 MWh in total, burnup of the first batch of fuel was 22.55%. Reactor core was made of 56 fuel channels. Activities related to construction of new and improvement of the existing equipment was continued in order to enable safe operation and successful utilization of the RA reactor. Exchange of the electronic tubes was continued in order to increase the stability of the reactor control and reactor protection systems. About 65% of tubes planned to be exchanged this year was done. Cooperation with the CEN Saclay, France related to construction of experimental loops VISA-1 and VISA-2 was continued as well as cooperation with Poland concerned with exchange of experts. The problem of lack of properly trained staff was nor solved. [Serbo-Croat] U izvestajnom periodu reaktor je uvek bio spreman za rad i odgovarao je zahtevima eksperimentatora. Reaktor je radio 408,5 casova na snagama od 50-5000 kW, odnosno ukupno 1985 MWh, pri cemu je ukupan utrosak prve sarze goriva bio 22,55%. Jezgro reaktora bilo je sacinjeno od 56 tehnoloskih kanala. Radi bezbednijeg pogona i sto uspesnije eksploatacije reaktora RA nastavljeno je sa radovima na realizaciji novih i poboljsanju postojecih uredjaja. U cilju povecanja stabilnosti elektronskih instrumenta u sistemu upravljanja i zastiti reaktora nastavljeno je sa zamenom elektronskih cevi, do sada je zamenjeno oko 65% radova predvidjenih za ovu godinu. Nastavljena je saradnja sa nuklearnim centrom u Saclay, Francuska na izradi petlje VISA-1 i VISA-2 i saradnja sa Poljskom u razmeni strucnjaka. Problem nedostatka kadrova nije resen.

  15. Professional Nuclear Materials Management; Gestion Industrielle des Matieres Nucleaires; Obrashchenie s yadernymi materialami na professional'nom urovne; Administracion Eficiente de Materiales Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Forcella, A. A.; O' Leary, W. J. [Allis-Chalmers Manufacturing Company, Bethesda, MD (United States)

    1966-02-15

    manipulacion y su transporte resulten economicos y sea facil sustituirlos. La programacion de las operaciones de fabricacion debera reducir al mfnimo las perdidas de ingresos debidas a un almacenamiento improductivo excesivamente largo de materiales de costo elevado. Para la etapa de su empleo en el reactor, sera necesario un programa adecuado de redistribucion de los elementos combustibles a fin de lograr el grado maximo de combustion del material fisionable. Ademas, habra que reducir en la medida de lo posible los paros improductivos del reactor debidos a la redistribucion del combustible, a las inspecciones, etc. Cuando el combustible haya llegado a un grado predeterminado de agotamiento, el administrador tendra que disponer la forma mas economica de regeneracion de los materiales fisionables y de recuperacion de los subproductos. La administracion de materiales nucleares es por consiguiente un factor esencial para la reduccion de los costos del ciclo del combustible y para lograr que el costo unitario de la energia producida resulte economicamente interesante. (author) [Russian] Daets ja opisanie ob{sup e}ma rabot na tipichnom jenergetichesko m reaktore v SShA. Poskol'ku jetot reaktor finansiruetsja chastnym kapitalom, odna iz osnovnyh objazannostej operatora sostoit v obespechenii sohrannosti kapitalovlozhenij i poluchenii opredelennoj pribyli. Vvidu bol'shoj sebestoimosti jadernyh materialov neobhodimo postojanno obespechivat' nadlezhashhuju bezopasnost' i uchet s cel'ju svedenija k minimumu poter', ne svjazannyh s obespe- cheniem bezopasnosti i ucheta jadernyh materialov. Neobhodimo vse tshhatel'no produmyvat' i planirovat' zaranee, chtoby izbegat' nenuzhnyh zatrat i rashodovanija kapitala ili snizhe- nija razmerov pribyli v rjade oblastej. Pojetomu administrator, vedajushhij jadernymi materialami, dolzhen zaranee uchityvat' vse nepredvidennye obstojatel'stva i osushhestvljat' postojannyj kontrol' nad otklonenijami ot standartov pri planirovanii vo vremja a

  16. Nuclear Reactor RA Safety Report, Vol. 13, Causes of possible accidents

    International Nuclear Information System (INIS)

    1986-11-01

    This volume includes the analysis of possible accidents on the RA research reaktor. Any unwanted action causing decrease of integrity of any of the reactor safety barriers is considered to be a reactor accident. Safety barriers are: fuel element cladding, reactor vessel, biogical shield, and reactor building. Reactor accidents can be classified in four categories: (1) accidents caused by reactivity changes; (2) accidents caused by mis function of the cooling system; (3) accidents caused by errors in fuel management and auxiliary systems; (4) accidents caused by natural or other external disasters. The analysis of possible causes of reactor accidents includes the analysis of possible impacts on the reactor itself and the environment [sr

  17. Experimental verification of reflector savings calculated by REDIR code using two-group method; Eksperimentalna provera dvogrupnog racunanja reflektorske ustede koriscenog u REDIR-u

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Martinc, R [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1968-12-15

    Radial buckling and reflector savings for heavy water reactor with 2% enriched uranium fuel were measured and calculated by the REDIR code. A comparison of the obtained values is presented in this paper dependent on the reactor lattice pitch and reflector thickness. Experimental results obtained for lattice pitch of 16 cm prove the validity of applying the REDIR code for power reactors. U radu je dato uporedjenje izmedju izmerenih i teorijski izracunatih vrednosti (prema programu REDIR) radijalnih baklinga i reflektorske ustede za teskovodni reaktorski sistem sa 2% obogacenim uranskim gorivom u zavisnosti od koraka resetke i debljine reflektora. Rezultati dobijeni eksperimentima pri koraku resetke od 16 cm potvdjuju ispravnost primene programa REDIR za energetske reaktore. (author)

  18. Vermikompos Sampah Kebun dengan Menggunakan Cacing Tanah Eudrilus eugeneae dan Eisenia fetida

    Directory of Open Access Journals (Sweden)

    Etik Rahmawati

    2016-04-01

    Full Text Available Durasi yang panjang diperlukan dalam pengomposan konvensional sampah organik yang memerlukan waktu selama 2-3 bulan. Pengurangan waktu pengomposan dapat dilakukan dengan digunakannya cacing sebagai dekomposer. Penelitian ini bertujuan untuk menentukan tingkat degradasi sampah kebun menggunakan proses vermikomposting dan menentukan pengaruh jenis cacing Eudrilus eugeneae dan Eisenia fetida. Empat reaktor berukuran 8 L digunakan dalam penelitian ini. Percobaan dilakukan secara duplo selama 60 hari. Parameter yang dianalisis pada penelitian ini adalah ammonia nitrogen (NH3-N, nitrat nitrogen (NO3-N, Total Kjeldahl Nitrogen (TKN, dan C/N. Hasil penelitian menunjukkan bahwa tingkat degradasi sampah kebun dengan pengolahan vermikomposting yang dapat dicapai adalah 64,94-72,52%. Produksi kompos yang lebih tinggi dengan penggunaan Eisenia fetida.

  19. WiN-Global 2007 Country Report for South Africa

    Energy Technology Data Exchange (ETDEWEB)

    Naidoo, Kameshni [South African National Nuclear Regulator, 17 Atlantic Road, Duynefontein, 07441 Cape Town (South Africa)

    2008-07-01

    South African nuclear landscape: - Koeberg Nuclear Power Station: Koeberg has been generating electricity for the past 19 years. Work began shortly after the contract with the French consortium was signed in 1976. In 1984 Koeberg started their commercial operation. Koeberg is located 30 km north of Cape Town. Comprises of two 900 MWe Pressurised Water Reactors. Produces 6.5% of South Africa's electricity needs. The reactor at Koeberg is cooled by cold water from the Atlantic Ocean. Low and intermediate level waste from Koeberg is transported by road in steel and concrete containers to a rural disposal site at Vaalputs, 600 km away in the Kalahari Desert. - PBMR company: The PBMR team is currently preparing for the building of a commercial scale power reactor project at Koeberg near Cape Town, where Africa's only nuclear power station is based, and a fuel plant at Pelindaba near Pretoria, where the pebble fuel will be manufactured. PBMR is a High Temperature Reactor (HTR) with a closed cycle, gas turbine power conversion system. Although it is not the only HTR currently being developed in the world, the South African project is on schedule to be the first commercial scale HTR in the power generation field. A steel pressure vessel holds the enriched uranium dioxide fuel encapsulated in graphite spheres. The system is cooled with helium and heat is converted into electricity through a turbine. - iThemba Labs: The iThemba Laboratory for Accelerator-Based Sciences is a group of multi-disciplinary research laboratories administered by the National Research Foundation. Based at two sites in Western Cape and Gauteng, these provide facilities for: Basic and applied research using particle beams, Particle radiotherapy for the treatment of cancer. The supply of accelerator-produced radioactive isotopes for nuclear medicine and research. iThemba Labs focuses on providing scientifically and medically useful radiation through the acceleration of charged particles using

  20. Study of indicators aggregation techniques for the selection of a new nuclear reactor for Mexico

    International Nuclear Information System (INIS)

    Barragan M, A.M.; Martin del Campo M, C.

    2007-01-01

    A study on several aggregation techniques that can be used as multi criteria analysis methods, like important part of the methodology developed for the selection of a nuclear reactor for Mexico is described. In an arbitrary way three reactors were selected to be compared, these they are the AP1000 (Advance Passive from 1000 MWe), the PBMR (Pebble Bed Modular Reactor) and the GT-MHR (Gas Turbine Modular Helium). The evaluation approaches were classified in three categories: Economic, Socio-political and of safety and environment. In each category they were defined the more important evaluation indicators and then it was built a matrix with those values of each reactor. The four studied aggregation methods are described: Normalization, Linear deliberation, Fuzzy Logic and AHP (Analytic Hierarchy Process). The well-known aggregation mechanisms are those that are obtained of the lineal deliberation and of the normalization, which have demonstrated to give good results before the simplicity of their use. The fuzzy logic has the advantage that it allows to manage qualitative and quantitative information simultaneously without the aggregation problems that are presented since in a conventional system the semantic pattern on that is based, it is provided by the theory of the diffuse groups that has demonstrated in other areas of the knowledge a better approach to the reality, when admitting that the nature has shades and that the decisions take in function of a wide range of possibilities and of approaches in contradictory occasions or in conflict, all equally worth. The Analytic Hierarchical Process (AHP) that consists in formalizing the intuitive understanding of a multi criteria complex problem, by means of the construction of a hierarchical model that allows the decision agent to structure the problem in visual form, giving him the form of a hierarchy of attributes (global objective of the problem, approaches and alternative). Finally, using the matrix of initiators

  1. STUDY ON DISCHARGE HEAT UTILIZATION OF 250 MWe PCMSR TURBINE SYSTEM FOR DESALINATION USING MODIFIED MED

    Directory of Open Access Journals (Sweden)

    Andang Widiharto

    2015-03-01

    Full Text Available PCMSR (Passive Compact Molten Salt Reactor is one type of Advanced Nuclear Reactors. The PCMSR has benefit charasteristics of very efficient fuel use, high safety charecteristic as well as high thermodinamics efficiency. This is due to its breeding capability, inherently safe characteristic and totally passive safety system. The PCMSR design consists of three module, i.e. reactor module, turbine module and fuel management module. Analysis in performed by parametric calculation of the turbine system to calculate the turbine system efficiency and the hat available for desalination. After that the mass and energi balance of desalination process are calculated to calculate the amount of distillate produced and the amount of feed sea water needed. The turbine module is designed to be operated at maximum temperature cycle of 1373 K (1200 0C and minimum temperature cycle of 333 K (60 0K. The parametric calculation shows that the optimum turbine pressure ratio is 4.3 that gives the conversion efficiency of 56 % for 4 stages turbine and 4 stages compressor and equiped with recuperator. In this optimum condition, the 250 MWe PCMSR turbine system produces 196 MWth of waste heat with the temperature of cooling fluid in the range from 327 K (54 0C to 368 K (92 0C. This waste heat can be utilized for desalination. By using MMED desalination system, this waste heat can be used to produce fresh water (distillate from sea water feed. The amount of the destillate produced is 48663 ton per day by using 15 distillation effects. The performance ratio value is 2.8727 kg/MJ by using 15 distillation effects. Keywords: PCMSR, discharged heat, MMED desalination   PCMSR (Passive Compact Molten Salt Reactor merupakan salah satu tipe dari Reaktor Nuklir Maju. PCMSR memiliki keuntungan berupa penggunaan bahan bakar yang sangat efisisien, sifat keselamatan tinggi dan sekaligus efisiensi termodinamika yang tinggi. Hal ini disebabkan oleh kemampuan pembiakan bahan bakar, sifat

  2. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  3. The Quality Standard 50-SG-Q8 Application On Research And Development Activities In P2TRR

    International Nuclear Information System (INIS)

    Nababan, N.; Rofei, A.

    2000-01-01

    Utilization of nuclear reactor installation should reference safety, Environment, and instalation requrement, and shoul incorporate user requirement. The requirement are adequately addressed in the IAEA standar No. 50-sg-Q8 for R and D work in nuclear instalation. A study to this standar has been carried out and it is found that there are three important factor of QA methodologi necessary to be aplied for R and D activities. They are management responsibilities, performance of the R and D works, and carrying out assesment over the R and D project. The R and D activities performing in Centre for Recearch Reaktor Technology Development (P2TRR) shoul meet the requirement in the R and D program assigned by Batan management, Such as Strastegy and Policy (jakarta), Strategic Plan (Restra), and Five-Year Plan. A study to this program also has been carried out and it is found that QA methodologies are necessary to be aplied, Especially during the data gathering activities, and its contribution to the safety function or as input to actual nuclear reactor installation or systems

  4. Localization of the Hot Spot in the Gap of Pebble Bed of Very High Temperature Gas Cooled Reactor(VHTGR)

    International Nuclear Information System (INIS)

    Lee, Sa Ya; Hong, Sung Je; Lee, Jae Young

    2010-01-01

    Pebble Bed Reactor(PBR) has been investigated intensively due to its benefits in management, but its complicated flow geometry requests reliable analytical methods. Hassan and Lee et al. have been made three dimensional computational methods. Hassan also measured local velocity fields with Particle Tracking Velocimetry(PTV), in small sized packed bed using liquid coolant, and Lee et al. measured flow field in the 2-dimensional wind tunnel with a hot wire system. In the present study, we develop the scaled up wind tunnel of pebble bed to use air as coolant in the same Reynolds number condition, as 21614, of the PBMR-250MWth. In order to measure the local surface temperature, the heating system and temperature measurement system were installed and heat transfer analogy was performed. The local surface temperature data shows that the predicted hot spots by Lee et al. at the top and bottom of the pebble by the velocity field measurement are reasonable, but the heat conduction is prior than contact effect at contact points

  5. Preliminary Cost Estimates for Nuclear Hydrogen Production: HTSE System

    International Nuclear Information System (INIS)

    Yang, K. J.; Lee, K. Y.; Lee, T. H.

    2008-01-01

    KAERI is now focusing on the research and development of the key technologies required for the design and realization of a nuclear hydrogen production system. As a preliminary study of cost estimates for nuclear hydrogen systems, the hydrogen production costs of the nuclear energy sources benchmarking GTMHR and PBMR are estimated in the necessary input data on a Korean specific basis. G4-ECONS was appropriately modified to calculate the cost for hydrogen production of HTSE (High Temperature Steam Electrolysis) process with VHTR (Very High Temperature nuclear Reactor) as a thermal energy source. The estimated costs presented in this paper show that hydrogen production by the VHTR could be competitive with current techniques of hydrogen production from fossil fuels if CO 2 capture and sequestration is required. Nuclear production of hydrogen would allow large-scale production of hydrogen at economic prices while avoiding the release of CO 2 . Nuclear production of hydrogen could thus become the enabling technology for the hydrogen economy. The major factors that would affect the cost of hydrogen were also discussed

  6. Three-Dimensional Analysis of the Hot-Spot Fuel Temperature in Pebble Bed and Prismatic Modular Reactors

    International Nuclear Information System (INIS)

    In, W. K.; Lee, S. W.; Lim, H. S.; Lee, W. J.

    2006-01-01

    High temperature gas-cooled reactors(HTGR) have been reviewed as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor(PBR) and a prismatic modular reactor(PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both a PBR and a PMR. The objective of this study is to predict the hot-spot fuel temperature distributions in a PBR and a PMR at a steady state. The computational fluid dynamics(CFD) code, CFX-10 is used to perform the three-dimensional analysis. The latest design data was used here based on the reference reactor designs, PBMR400 and GTMHR60

  7. Perhitungan Kebutuhan Cooling Tower Pada Rancang Bangun Untai Uji Sistem Kendali Reaktor Riset

    OpenAIRE

    Awwaluddin, Muhammad; Santosa, Puji; Suwardiyono, Suwardiyono

    2012-01-01

    CALCULATION OF THE NEED FOR COOLING TOWER ON DESIGN OF STRAND TEST RESEARCH REACTOR CONTROL SYSTEM. Cooling tower on the strand test engineering research reactor control system functioning as a heat transfer medium from the heat exchanger to air. To get the transfer of heat or cooling is maximal then the determination of cooling tower needs to be precise. Cooling tower is expected to accept and release heat at 1.191 kw from the heat exchanger. To support these needs will require the calculati...

  8. RA reactor Action plan for 1978 - Annex VI; Prilog VI - Plan rada reaktora RA za 1978. godinu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-12-15

    This annex include a detailed monthly action plan of the RA reactor for 1978. It was planned that the reactor will operate at nominal and lower power levels for 158 days. The plan is made with the aim to achieve same annual neutron flux as in 1977, which is somewhat higher then the annual production in the period from 1968-1975. [Serbo-Croat] Ovaj prilog sadrzi detaljan plan rada reaktora RA za 1978. godinu. Planirano je da reaktor radi na nominalnoj snazi of 6,5 MW i na manjim snagama ukupno 158 dana. Plan je napravljen tako da se ostvari godisnja proizvodnja neutronskog fluksa u obimu koji je relaizovan 1977. godine sto je nesto vise od godisnje proizvodnje u periodu 1968-1975.

  9. Definition of task (Phase II) B, 'Leak testing'; Definicija zadatka (II faza) B, 'Ispitivanje hermeticnosti'

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M; Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-01-15

    In order to ensure safe Ra reactor operation and fulfilling the conditions for performing irradiation experiments it was necessary to test and verify the leak tightness of the sample containers and thermocouples, as well as the irradiation capsule before it was placed in the VISA-2 channel. Leak testing of VISA-2 channel with capsules and thermocouples was done before and after it was built-in the reactor. [Serbo-Croat] U cilju bezbednosti i sigurnosti rada reaktora RA i ispunjavanja uslova ozracivanja izvrsena je provera hermeticnosti kenera sa uzorcima i termoparovima, kao i hermeticnost kapsule pre ugradjivanja kapsule u kanal VISA-2. Izvrseno je i ispitivanje hermeticnosti kanala VISA-2 sa kapsulama i termoparovima pre ugradjivanja u reaktor i posle montaze.

  10. Validation of CATHARE for gas-cooled reactors

    International Nuclear Information System (INIS)

    Fabrice Bentivoglio; Ola Widlund; Manuel Saez

    2005-01-01

    Full text of publication follows: Extensively validated and qualified for light-water reactor safety studies, the thermo-hydraulics code CATHARE has been adapted to deal also with gas-cooled reactor applications. In order to validate the code for these novel applications, CEA (Commissariat a l'Energie Atomique) has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE is being validated against existing experimental data, in particular from the German power plant Oberhausen II and the South African Pebble-Bed Micro Model (PBMM). Oberhausen II, operated by the German utility EVO, is a 50 MW(e) direct-cycle Helium turbine plant. The power source is a gas burner rather than a nuclear reactor core, but the power conversion system resembles those of the GFR (Gas-cooled Fast Reactor) and other high-temperature reactor concepts. Oberhausen II was operated for more than 100 000 hours between 1974 and 1988. Design specifications, drawings and experimental data have been obtained through the European HTR project, offering a unique opportunity to validate CATHARE on a large-scale Brayton cycle. Available measurements of temperatures, pressures and mass flows throughout the circuit have allowed a very comprehensive thermohydraulic description of the plant, in steady-state conditions as well as during transients. The Pebble-Bed Micro Model (PBMM) is a small-scale model conceived to demonstrate the operability and control strategies of the South African PBMR concept. The model uses Nitrogen instead of Helium, and an electrical heater with a maximum rating of 420 kW. As the full-scale PBMR, the PBMM loop features three turbines and two compressors on the primary circuit, located on three separate shafts. The generator, however, is modelled by a third compressor on a separate circuit, with a

  11. Microwave integrated circuit radiometer front-ends for the Push Broom Microwave Radiometer

    Science.gov (United States)

    Harrington, R. F.; Hearn, C. P.

    1982-01-01

    Microwave integrated circuit front-ends for the L-band, S-band and C-band stepped frequency null-balanced noise-injection Dicke-switched radiometer to be installed in the NASA Langley airborne prototype Push Broom Microwave Radiometer (PBMR) are described. These front-ends were developed for the fixed frequency of 1.413 GHz and the variable frequencies of 1.8-2.8 GHz and 3.8-5.8 GHz. Measurements of the noise temperature of these units were made at 55.8 C, and the results of these tests are given. While the overall performance was reasonable, improvements need to be made in circuit losses and noise temperatures, which in the case of the C-band were from 1000 to 1850 K instead of the 500 K specified. Further development of the prototypes is underway to improve performance and extend the frequency range.

  12. Obese Japanese adults with type 2 diabetes have higher basal metabolic rates than non-diabetic adults.

    Science.gov (United States)

    Miyake, Rieko; Ohkawara, Kazunori; Ishikawa-Takata, Kazuko; Morita, Akemi; Watanabe, Shaw; Tanaka, Shigeho

    2011-01-01

    Several cross-sectional studies in Pima Indians and Caucasians have indicated that obese individuals with type 2 diabetes have a higher basal metabolic rate (BMR) than healthy, obese individuals. However, no study has investigated this comparison in Japanese subjects, who are known to be susceptible to type 2 diabetes due to genetic characteristics. Thirty obese Japanese adults with pre-type 2 diabetes (n=7) or type 2 diabetes (n=13) or without diabetes (n=10) participated in this study. BMR was measured using indirect calorimetry. The relationships between residual BMR (calculated as measured BMR minus BMR adjusted for fat-free mass, fat mass, age, and sex) and biomarkers including fasting glucose, glycosylated hemoglobin (HbA(1c)), fasting insulin, homeostasis model assessment of insulin resistance (HOMA-R), triglycerides, and free fatty acids were examined using Pearson's correlation. BMR in diabetic subjects adjusted for fat-free mass, fat mass, age, and sex was 7.1% higher than in non-diabetic subjects. BMR in diabetic subjects was also significantly (pBMR and fasting glucose (r=0.391, p=0.032). These results indicate that in the Japanese population, obese subjects with type 2 diabetes have higher BMR compared with obese non-diabetic subjects. The fasting glucose level may contribute to these differences.

  13. Thermo-economic performance of HTGR Brayton power cycles

    International Nuclear Information System (INIS)

    Linares, J. L.; Herranz, L. E.; Moratilla, B. Y.; Fernandez-Perez, A.

    2008-01-01

    High temperature reached in High and Very High Temperature Reactors (VHTRs) results in thermal efficiencies substantially higher than those of actual nuclear power plants. A number of studies mainly driven by achieving optimum thermal performance have explored several layout. However, economic assessments of cycle power configurations for innovative systems, although necessarily uncertain at this time, may bring valuable information in relative terms concerning power cycle optimization. This paper investigates the thermal and economic performance direct Brayton cycles. Based on the available parameters and settings of different designs of HTGR power plants (GTHTR-300 and PBMR) and using the first and second laws of thermodynamics, the effects of compressor inter-cooling and of the compressor-turbine arrangement (i.e., single vs. multiple axes) on thermal efficiency have been estimated. The economic analysis has been based on the El-Sayed methodology and on the indirect derivation of the reactor capital investment. The results of the study suggest that a 1-axis inter-cooled power cycle has a similar thermal performance to the 3-axes one (around 50%) and, what's more, it is substantially less taxed. A sensitivity study allowed assessing the potential impact of optimizing several variables on cycle performance. Further than that, the cycle components costs have been estimated and compared. (authors)

  14. KETAHANAN KOROSI PADUAN Al-Mg 5052 DI DALAM AIR PENDINGIN NETRAL MENGANDUNG KLORIDA

    Directory of Open Access Journals (Sweden)

    Dicky Tri Jatmiko

    2015-07-01

    Full Text Available KETAHANAN KOROSI PADUAN Al-Mg 5052 DI DALAM AIR PENDINGIN NETRAL MENGANDUNG KLORIDA. Paduan Al-Mg 5052 adalah material yang biasa digunakan untuk kelongsong elemen bakar nuklir karena serapan fluks netronnya rendah dan tahan korosi di dalam air demineralisasi pada kondisi operasi reaktor. Makalah ini difokuskan untuk mengetahui ketahanan korosi paduan Al-Mg 5052 di dalam air dengan pH netral dan mengandung klorida sebagai pengganti air demineralisasi pendingin primer Reaktor Serba Guna GA Siwabessy (RSG-GAS. Penelitian mencakup pengukuran laju korosi menggunakan metode Tafel, prediksi mekanisme korosi menggunakan metode voltametri siklik dan analisa produk korosi dengan metode difraksi sinar X. Percobaan dilakukan dengan variasi temperatur 30°C, 35°C, 40°C, dan 45°C, serta variasi konsentrasi larutan natrium  klorida 0,05 M, 0,25 M, dan 0,5 M. Hasil penelitian ini menunjukkan bahwa paduan Al-Mg 5052 terkorosi dengan kategori “dapat diabaikan” hingga “sedang” dalam larutan natrium klorida menjadi produk yang larut dalam air pada satu tahap reaksi oksidasi irreversible.   CORROSION RESISTANCE OF Al-Mg ALLOY 5052 IN CHLORIDE CONTAINING NEUTRAL COOLING WATER. Al-Mg alloy 5052 is a material used as nuclear fuel element cladding due to its low neutron flux absorption and high corrosion resistance in demineralized water. This research is focused to know of the corrosion resistance of Al-Mg alloy 5052 in chloride containing neutral water used as demineralized primary cooling water substitute in GA Siwabessy Multi Purpose Reactor (RSG-GAS. This research covers the corrosion rate measurement using the Tafel method, corrosion process prediction using cyclic voltammetry method and corrosion product analysis using X-Ray Diffraction method. The experiments are carried out at temperature variation of 30°C, 35°C, 40°C and 45°C, as well as sodium chloride concentration of 0.05 M, 0.25 M and 0.5 M. The research results show that Al-Mg alloy 5052

  15. Commissioning Experience from the Agesta Nuclear Power Plant; Experience acquise lors des essais de mise en service de la centrale nucleaire d'Agesta; Opyt po vvedeniyu v ehkspluatatsiyu yadernoj ehnergeticheskoj ustanovki Agesta; Experiencia adquirida con la puesta en marcha de la central nucleoelectrica de Agesta

    Energy Technology Data Exchange (ETDEWEB)

    Rydell, N. [Aagesta Kraftvarmewerk, Farsta (Sweden)

    1963-10-15

    The Agesta Nuclear Power Plant is a pressurized heavy water reactor of the pressure vessel type, fuelled with natural uranium. It was commissioned with light water from December 1962 to May 1963. Observations of a more general interest were made during this commissioning essentially on the following topics; (a) cleanliness of primary circuit (b) valve operation (c) pressurization of the primary circuit (d) water leakage (e) refuelling machinery (f) containment testing. (author) [French] Il s'agit d'un reacteur a uranium naturel et a eau lourde pressurisee, du type a caisson sous pression. Les essais de mise en service ont ete faits avec de l'eau ordinaire, de decembre 1962 a mai 1963. La mise en service a permis de faire des observations d'interet general sur les sujets suivants: a) non-contamination du circuit primaire; b) fonctionnement des vannes; c) maintien sous pression du circuit primaire; d) fuites d'eau; e) appareils de chargement du combustible; f) essais d'isolement. (author) [Spanish] La central nucleoelectrica de Agesta posee un reactor de agua pesada del tipo de recipiente de presion, con combustible de uranio natural. Se mantuvo en funcionamiento con agua ligera entre diciembre de 1962 y mayo de 1963. Durante esta prueba, se efectuaron observaciones de interes mas general, relacionadas esencialmente con las siguientes cuestiones: a) limpieza del circuito primario; b) funcionamiento de las valvulas; c) presion del circuito primario; d) perdidas de agua; e) dosposiciones de reposicion del Combustible; f) ensayos de confinamiento. (author) [Russian] Yadernaya ehnergeticheskaya ustanovka Agesta predstavlyaet soboj tyazhelovodnyj reaktor pod davleniem, ispol'zuyushchij prirodnyj uran v kachestve topliva. Reaktor byl vveden v ehkspluatatsiyu na obychnoj vode v period s dekabrya 1962 goda po maj 1963 goda. Zamechaniya bolee obshchego kharaktera byli sdelany vo vremya ehkspluatatsii v osnovnom po sledukhshchim temam: a) chistota pervichnogo kontura; b

  16. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    Science.gov (United States)

    Karim, Julia Abdul

    2008-05-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained.

  17. Core Calculation of 1 MWatt PUSPATI TRIGA Reactor (RTP) using Monte Carlo MVP Code System

    International Nuclear Information System (INIS)

    Karim, Julia Abdul

    2008-01-01

    The Monte Carlo MVP code system was adopted for the Reaktor TRIGA PUSAPTI (RTP) core calculation. The code was developed by a group of researcher of Japan Atomic Energy Agency (JAEA) first in 1994. MVP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation and able to estimate an accurate simulation problems. The code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique faster than conventional method as compared to the conventional scalar method. This code could achieve higher computational speed by several factors on the vector super-computer. In this calculation, RTP core was modeled as close as possible to the real core and results of keff flux, fission densities and others were obtained

  18. Fuel element transfer cask modelling using MCNP technique

    International Nuclear Information System (INIS)

    Rosli Darmawan

    2009-01-01

    Full text: After operating for more than 25 years, some of the Reaktor TRIGA PUSPATI (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement. (author)

  19. Fuel Element Transfer Cask Modelling Using MCNP Technique

    International Nuclear Information System (INIS)

    Darmawan, Rosli; Topah, Budiman Naim

    2010-01-01

    After operating for more than 25 years, some of the Reaktor TRIGA Puspati (RTP) fuel elements would have been depleted. A few addition and fuel reconfiguration exercises have to be conducted in order to maintain RTP capacity. Presently, RTP spent fuels are stored at the storage area inside RTP tank. The need to transfer the fuel element outside of RTP tank may be prevalence in the near future. The preparation shall be started from now. A fuel element transfer cask has been designed according to the recommendation by the fuel manufacturer and experience of other countries. A modelling using MCNP code has been conducted to analyse the design. The result shows that the design of transfer cask fuel element is safe for handling outside the RTP tank according to recent regulatory requirement.

  20. The Ellweiler uranium plant - a demolition and recycling project

    International Nuclear Information System (INIS)

    Mika, S.; Rohr, T.; Seehars, R.; Feser, A.

    1999-01-01

    The uranium plant at Ellweiler, district of Birkenfeld, was used for the production and storage of uranium concentrates. The owner of the Ellweiler uranium plant (UAE), Gewerkschaft Brunhilde GmbH, ceased processing uranium ore and recycling in 1989 and has been in liquidation since September 1991. The State of Rhineland-Palatinate, had safety measures adopted in a first step, getting the plant into a safe state by former plant personnel. The entire plant was demolished in a second step. The contract for demolishing the former uranium plant was awarded to ABB Reaktor as the general contractor in August 1996. Demolition work was carried out between April 1997 and May 1999. A total of approx. 7900 Mg of material was disposed of. At present, recultivation measures are being carried out. (orig.) [de

  1. Progress in the development of uranium silicide (U3Si2) fuel at BATAN

    International Nuclear Information System (INIS)

    Suripto, A.; Soentono, S.

    1995-01-01

    After successful fabrication of two full-size prototype fuel elements containing ∼3.0 gU/cm 3 in the form of U 3 Si 2 -Al dispersion now undergoing irradiation in the Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) core since 1990, further development in U 3 Si 2 -A2 dispersion fuel element manufacturing has been pursued, whose progress in discussed in this paper, with a special attention on the use of much higher-loading aimed at obtaining a better understanding on the influence of higher-loading on fuel core and plate manufacturing and quality. At present, high-loading U 3 Si 2 -AI dispersion miniplates are being manufactured for preparing some mini-fuel elements to be test-irradiated in the new MTR in-pile loop of the RSG-GAS. (author)

  2. Experimental operation of the RA reactor with 4 fuel channels containing 80% enriched dispersion fuel - Operational Report; Radni izvestaj - Eksperimentalna kampanja reaktora RA sa 4 kanala sa 80% obogacenim disperzionim gorivom

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Milosevic, M; Cupac, S; Kozomara, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    Start of utilization of the new 80% enriched dispersion nuclear fuel is underway in the RA reactor core. Both economic and technical analyses were in favor of introducing the new fuel elements gradually into the RA reactor core. Thus overall theoretical and experimental analyses as well as other preparations are directed to transition regime based on gradual introducing of new fuel into the core, i.e. reactor core with two types of fuel. The objective of these analyses and preparation is establishment of conditions for safe reactor operation during transition period. The analyses and preparations are almost completed. The experimental data about fuel burnup during a time period of operation at nominal power i.e. daily decrease of excess reactivity is missing. This data is needed for planning the refueling (quantity of fresh fuel and frequency of refueling) during the transient period. This data can be obtained only by normal operation of the reactor during a period of time significantly longer than the period of attaining equilibrium poisoning, as time between two D{sub 2}O condensate overflows into the RA reactor core. Thus a ten day experimental campaign was planned to be done in December 1976. This report presents the most important results of safety analyses and preparation which show that, during this experimental period, the reactor operation is absolutely safe taking into account the most important parameters influencing reactor safety, as reactivity, thermal and temperature limits for fuel and the reactor, etc. Data to be obtained during this experimental campaign are significant because they would enable definition of future supply of fresh fuel during the transition period. [Serbo-Croat] Predstoji pocetak koriscenja novog 80% obogacenog uranskog disperzionog goriva u reaktoru RA. Ekonomske i tehnicke analize dale su prednost postupku postepenog uvodjenja novog goriva u reaktor RA. Prema tome, obimne teorijske i eksperimentalne analize i druge pripreme

  3. PENGOLAHAN LIMBAH CAIR INDUSTRI FARMASI FORMULASI DENGAN METODE ANAEROB-AEROB DAN ANAEROB-KOAGULASI

    Directory of Open Access Journals (Sweden)

    Farida Crisnaningtyas

    2016-05-01

    Full Text Available Studi ini membahas mengenai pengolahan limbah cair industri farmasi dalam skala laboratorium dengan menggunakan konsep anaerob-kimia-fisika dan anaerob-aerob. Proses anaerob dilakukan dengan menggunakan reaktor Upflow Anaerobic Sludge Bed reactor (UASBr pada kisaran OLR (Organic Loading Rate 0,5 – 2 kg COD/m3hari, yang didahului dengan proses aklimatisasi menggunakan substrat gula. Proses anaerob mampu memberikan efisiensi penurunan COD hingga 74%. Keluaran dari proses anaerob diolah lebih lanjut dengan menggunakan dua opsi proses: (1 fisika-kimia, dan (2 aerob. Koagulan alumunium sulfat dan flokulan kationik memberikan efisiensi penurunan COD tertinggi (73% pada kecepatan putaran masing-masing 100 rpm dan 40 rpm. Uji coba aerob dilakukan pada kisaran MLSS antara 4000-5000 mg/L dan mampu memberikan efisiensi penurunan COD hingga 97%. Hasil uji coba menunjukkan bahwa efisiensi penurunan COD total yang dapat dicapai dengan menggunakan teknologi anaerob-aerob adalah 97%, sedangkan kombinasi anaerob-koagulasi-flokulasi hanya mampu menurunkan COD total sebesar 72,53%. Berdasarkan hasil tersebut, kombinasi proses anaerob-aerob merupakan teknologi yang potensial untuk diaplikasikan dalam sistem pengolahan limbah cair industri farmasi. 

  4. Shielding design of highly activated sample storage at reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Naim Syauqi Hamzah; Julia Abdul Karim; Mohamad Hairie Rabir; Muhd Husamuddin Abdul Khalil; Mohd Amin Sharifuldin Salleh

    2010-01-01

    Radiation protection has always been one of the most important things considered in Reaktor Triga PUSPATI (RTP) management. Currently, demands on sample activation were increased from variety of applicant in different research field area. Radiological hazard may occur if the samples evaluation done were misjudge or miscalculated. At present, there is no appropriate storage for highly activated samples. For that purpose, special irradiated samples storage box should be provided in order to segregate highly activated samples that produce high dose level and typical activated samples that produce lower dose level (1 - 2 mR/ hr). In this study, thickness required by common shielding material such as lead and concrete to reduce highly activated radiotracer sample (potassium bromide) with initial exposure dose of 5 R/ hr to background level (0.05 mR/ hr) were determined. Analyses were done using several methods including conventional shielding equation, half value layer calculation and Micro shield computer code. Design of new irradiated samples storage box for RTP that capable to contain high level gamma radioactivity were then proposed. (author)

  5. Investigations of coal ignition in a short-range flame burner using optical measuring systems; Untersuchungen zur Kohlezuendung am Flachflammenbrenner unter Verwendung optischer Messtechnik

    Energy Technology Data Exchange (ETDEWEB)

    Hackert, G.; Kremer, H.; Wirtz, S. [Bochum Univ. (Germany). Lehrstuhl fuer Energieanlagentechnik

    1999-09-01

    The short-range flame burner and the KOALA reactor of DMT are experimental facilities for realistic simulation of coal conversion processes at high temperatures and pressures in atmospheric conditions. The TOSCA system enable measurements of temperatures, sizes, shapes and velocities of the fuel particles, which serve as a basis for a three-dimensional simulation model of coal combustion. In the future, further parameter studies will deepen the present knowledge of coal dust combustion under pressure and enable optimisation of the numerical models for simulation of industrial-scale systems for coal dust combustion under pressure. [Deutsch] Mit dem Flachflammenbrenner und dem KOALA-Reaktor der DMT stehen Versuchsapparaturen zur Verfuegung, mit deren Hilfe die Kohleumwandlungsprozesse bei hohen Temperaturen unter Druck und unter atmosphaerischen Bedingungen realistisch wiedergegeben werden. Das TOSCA-System erlaubt dabei die Bestimmung von Temperaturen, Groessen, Formen und Geschwindigkeiten der Brennstoffpartikel. Diese Daten liefern die Grundlage fuer die Erstellung eines dreidimensionalen Simulationsmodells zur Modellierung der Kohleverbrennung. In Zukunft werden weitere Parameterstudien das Verstaendnis der Kohlenstaubdruckverbrennung vertiefen und ein Optimierung der numerischen Modelle ermoeglichen, so dass die Simulation grosstechnischer Kohlenstaubdruckverbrennungsanlagen realisiert werden kann. (orig.)

  6. CFD applications in the Pebble Bed Modular Reactor Project: A decade of progress

    International Nuclear Information System (INIS)

    Janse van Rensburg, J.J.; Kleingeld, M.

    2011-01-01

    Highlights: → This paper evaluates the evolution of Gas Cooled Reactor CFD analysis over the last decade. → It discusses the influence of advances in hardware and software on the evolution of capabilities. → The advances in mesh generation and the physics that can be included is also discussed. → The focus was on the capabilities rather than improving the assumptions and correlations. - Abstract: Of all the systems and components that have to be designed for a nuclear plant, the Reactor Unit is the most significant since it is at the very heart of the plant. At Pebble Bed Modular Reactor (Pty) Ltd. (PBMR), the design of the Reactor Unit is conducted with the aid of extensive analysis work. Due to the rapid computational improvements, the analysis capabilities have had to evolve rather significantly over the last decade. This paper evaluates the evolution of RU Computational Fluid Dynamics (CFD) analysis in particular and presents a historical timeline of the analyses conducted at PBMR. The influence of advances in the hardware and software applications on the evolution of the analysis capabilities is also discussed. When evaluating the evolution of analysis, it is important to look not only at the advances in mesh generation and the representation of the geometry, but also at the improvements regarding the physics that were included in the models. The discussion evaluates the improvements from the pre-conceptual analyses, the concept design, the basic design and finally, the detail design. It is however important to note that the focus of this research was on establishing a methodology for the integrated CFD analysis of High Temperature Reactors. It is recognized however that results from this research can currently only be used to investigate and understand trends and behaviors rather than absolute values. It was therefore required to also launch an extensive V and V program of which the focus was to verify the approach and validate the methodology that

  7. Metaphysics methods development for high temperature gas cooled reactor analysis

    International Nuclear Information System (INIS)

    Seker, V.; Downar, T. J.

    2007-01-01

    Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology road map. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark

  8. A Review on the VHTR PIRT Development Status of Both Regulatory Authority and Licensee

    International Nuclear Information System (INIS)

    Hwang, Su Hyun; Jeon, Seong Su; Hong, Soon Joon; Lee, Byung Chul; Huh, Chang Wook; Jin, Chang Yong; Kim, Kyun Tae

    2011-01-01

    The VHTR (Very High Temperature Reactor) is defined as a helium-cooled, graphite moderated reactor with a core outlet temperature in excess of 900 .deg. C and a long-term goal of achieving an outlet temperature of 1000 .deg. C. The VHTR is suited for a broad range of applications, including the production of hydrogen and electricity. The PIRT (Phenomena Identification and Ranking Table) provides a structured means of identifying and analyzing a wide variety of off-normal sequences that potentially challenge the viability of complex technological systems. As applied to VHTR, the PIRT is used to identify a spectrum of safety-related sequences or phenomena that could affect those systems, and to rank order those sequences on the basis of their frequencies, their potential consequences, and state of knowledge related to associate phenomena. It is to be used as an early screening tool to identify, categorize, and characterize phenomena and issues that are potentially important to risk and safety of VHTR. Since a specific design has not yet been selected for the choice of the US VHTR (NGNP), it was decided early on to focus on a generic plant and reactor design with broadly typical features. Both a generic Pebble Bed Reactor (PBR) design and a generic Prismatic Modular Reactor (PMR) design were selected as the reference plant for KAERI and ANL PIRT. The generic PBR design selected is a version of the 400 MWt South African PBMR design. The generic PMR design selected is a version of the 600 MWt GT-MHR. The reference plant of NRC PIRT is assumed to be a modular high temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GT-MHR) version (a prismatic-core modular reactor- PMR) or a pebble bed modular reactor (PBMR) version (a pebble bed reactor-PBR) design, with either a director indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. The difference of VHTR PIRT

  9. Nový úkol hybridních reaktorů – likvidace jaderného odpadu

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2009-01-01

    Roč. 57, č. 11 (2009), s. 24-a33 ISSN 0040-1064 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * ITER * hybrid * nuclear waste * LIFE * FFTS * CFNS * SXD * FDF Subject RIV: BL - Plasma and Gas Discharge Physics

  10. Research reactor 'A' 6.5/10 MW; Istrazivacki reaktor 'A' 6,5/10 MW

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1961-07-01

    This booklet includes a short description of the RA research reactor with basic properties of its components: control and safety system, heavy water system, technical water cooling system, heavy water distillation system, cover gas system, dosimetry control system, power supply system. It is used for fundamental research in reactor and nuclear physics, isotope production, materials testing.

  11. Use of plutonium and minor actinides as fuel in high temperature pebble bed reactors for waste minimization

    International Nuclear Information System (INIS)

    Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenther

    2009-01-01

    Energy production by nuclear fission gives rise to longlived radionuclides, such as plutonium and americium. The ''PuMA'' (Plutonium and Minor Actinides Waste Management) research project within the 6th Framework Program of the European Union serves to minimize waste arisings and transmute plutonium and minor actinides from spent LWR fuel elements by means of modular high-temperature reactors (HTR). Coating the fuel, which consists of kernels approx. 250 μm in radius and surrounded by graphite as the moderator material, allows very high operating and accident temperatures and very high burnups. One point examined is whether the inherent safety characteristics known for uranium oxide also exist for (PuO 2 + MAO 2 ) fuel. On the basis of a reference reactor similar to the South African PBMR-400, various loading strategies at maximum burnup are considered with a view to the inherent safety of the HTR. (orig.)

  12. TORT-TD/ATTICA3D: a coupled neutron transport and thermal hydraulics code system for 3-D transient analysis of gas cooled high temperature reactors

    International Nuclear Information System (INIS)

    Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.

    2011-01-01

    Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (author)

  13. Komposit Nano TiO2 Dengan PCC, Zeolit atau Karbon Aktif Untuk Menurunkan Total Krom dan Zat Organik Pada Air Limbah Industri Penyamakan Kulit

    Directory of Open Access Journals (Sweden)

    Bumiarto Nugroho Jati

    2012-04-01

    Full Text Available Telah dilakukan penelitian untuk menurunkan total krom dan zat organik pada limbah industri penyamakan kulit dengan menggunakan nano TiO2 yang dikompositkan dengan adsorben karbon aktif, zeolit, dan precipitated calcium carbonate (PCC dalam suatu reaktor fotokatalitik yang disusun secara batch dan dilengkapi dengan 6 buah lampu UV dan magnetic stirrer. Penurunan kadar krom total diukur dengan menggunakan Atomic Absorption Spectro-photometer (AAS dan penurunan zat organik dianalisa dengan menggunakan titrasi permanganatometri. Hasil penelitian menunjukkan pengolahan terbaik untuk penurunan kadar krom total adalah dengan menggunakan komposit TiO2:PCC = 8:2 yang dapat menurunkan total krom hampir 100% pada menit ke-170 dengan konsentrasi awal 214,35 mg/L. Untuk penurunan kadar zat organik, pengolahan terbaik dengan menggunakan komposit TiO2:PCC = 9:1 yang dapat menurunkan kadar zat organik hingga 100% pada menit ke-180. 

  14. Activities of the Service for maintenance of the RA reactor electronic equipment in 1979 - Report - Annex IV; Prilog IV Rad sluzbe za odrzavanje elektronske opreme reaktora RA u 1979. godini - Izvestaj -

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1979-12-15

    Within the organizational structure of the RA reactor staff, the Service for instrumentation maintenance has the following tasks: maintenance of the existing electronic equipment; participating in experiments planning and preparation of electronic equipment; purchasing new equipment, spare parts and components; construction of new equipment for internal needs; implementation of new equipment. Basic instrumentation of the reactor facility includes: control and protection system, and dosimetry system. [Serbo-Croat] U okviru organizacione seme OOUR nuklearni reaktor RA sluzba za odrzavanje instrumentacije postrojenja ima sledece zadatke: odrzavanje postojece elektronske opreme; ucesce u planiranju eksperimenata i priprema elektronske opreme; nabavka nove opreme, rezervnih delova i komponenti; rad na izradi elektronske opreme i uredjaja za sopstvene potrebe i izrada projekata u domenu reaktorske opreme i ucesce u ugradnji nove opreme. Osnovna instrumentacija postrojenja obuhvata: sistem za upravljanje i zastitu, sistem za tehnolosku kontrolu i sistem za dozimetrijsku kontrolu.

  15. Pengaruh Jenis Bahan pada Proses Pirolisis Sampah Organik menjadi Bio-Oil sebagai Sumber Energi Terbarukan

    Directory of Open Access Journals (Sweden)

    M. Sigit Cahyono

    2013-06-01

    Full Text Available Sampah organik merupakan potensi sumber energi yang melimpah di Indonesia. Sampah organik berupa daun dan ranting kering bisa dikonversi menjadi bahan bakar berupa bio-oil melalui proses fast pirolisis. Tujuan dari penelitian ini adalah untuk mengetahui pengaruh jenis bahan terhadap rendemen dan nilai kalor bio-oil yang dihasilkan dari proses pirolisis sampah organik. Bahan baku berupa daun dan ranting kering campuran tanaman angsana, mahoni dan mangga dengan komposisi daun bervariasi 0%, 50%, dan 100%, dipotong-potong dengan ukuran maksimal 10 cm. Kemudian bahan baku tersebut dipanaskan di dalam reaktor pirolisis pada suhu 500 C selama 1 jam. Hasil penelitian menunjukkan bahwa nilai kalor tertinggi (5175,35 J/g dan rendemen tertinggi (24,5% didapatkan pada bio-oil yang dihasilkan dari pirolisis ranting 100%. Kata kunci: Sampah Organik, Bio-oil, Pirolisis, Rendemen, Nilai Kalor

  16. MASYARAKAT MADURA DAN RENCANA PEMBANGUNAN PLTN; PERSPEKTIF TEOLOGI

    Directory of Open Access Journals (Sweden)

    Abd. A'la

    2012-05-01

    Full Text Available Abd A’la   (Staf pengajar pada program Pascasarjana IAIN Sunan Ampel dan Asiten Direktur Bidang Akademik pada lembaga yang sama       Abstrak : Setiap pemanfaatan energi selalu memiliki resiko environmental. Resiko dalam dunia ilmu dapat ditolelir jika hal itu telah terkalkulasikan sehingga merupakan caculable risk. Sementara untuk energi nuklir, sampai saat ini, belum dapat dilakukan terutama terkait dengan teknologi pengamannya sehingga Fitjof Capra menyebutnya sebagai rasionalitas yang tidak bertanggung jawab. Masyarakat Madura merupakan komunitas Sunny dan dalam merespon setiap perkembangan selalu berjangkar pada prinsip-prinsip Sunny yang dicirikan dengan kehati-hatian dan moderasi. Terkait dengan rencana PLTN madura, masalah data tentang kebutuhan hal itu harus segera diperoleh untuk menjawab perlu tidaknya PLTN di Madura. Kekhwatiran bahwa kebutuan itu hanya merupakan rekayasa perusahaan reaktor nuklir patut di pertimbangkan karena memang sangat beralasan Kata Kunci : masyarakat madura, nuklir, PLTN

  17. ANALISA THERMOGRAVIMETRY PADA PIROLISIS LIMBAH PERTANIAN

    Directory of Open Access Journals (Sweden)

    Bagus Setiawan

    2016-01-01

    Full Text Available Penelitian ini bertujuan untuk menemukan karakterisasi degradasi termal dari limbah pertanian untuk dijadikan suatu bahan bakar padat alternatif. Penelitian diawali dengan tahap pengumpulan bahan yang dilanjutkan penyeragaman ukuran sampel uji hingga berukuran 20 mesh. Setelah itu masing-masing sampel dikeringkan hingga kadar air maksimal 12 %. Sebelum Sampel seberat 20 gram diuji pirolisis dengan menempatkan sampel dalam reaktor yang telah dialiri nitrogen dengan laju 100 ml/menit. Sampel diuji dengan kondisi heating rate 15 oC/menit, temperatur akhir 600 oC dan holding time 10 menit. Data yang didapat berupa penurunan massa dan perubahan temperatur dicatat dalam laptop dengan menggunakan software RS-Key, Ms Excel dan Adam.NET Utility. Dari penelitian yang telah dilakukan, maka dapat disimpulkan campuran serbuk gergaji dan jerami memiliki temperatur pirolisis paling rendah, sementara campuran sekam padi dan kulit singkong memiliki massa arang paling banyak.

  18. [Study of Basal metabolic rate of 81 young adults aged 20-29 years old in Changsha].

    Science.gov (United States)

    Zhou, X; Mao, D Q; Luo, J Y; Wu, J H; Zhuo, Q; Li, Y M

    2017-07-06

    Objective: To determine the basal metabolic rate (BMR) of young adults aged between 20-29 years old in Changsha. Methods: We recruited volunteers to join in our research project from April to May, 2015. All recruited volunteers must meet the inclusion criteria: aged 20-29 years old, height between 164-180 centimeters in males and 154-167 centimeters in females, in good health condition, and with no habit of regular physical exercise in last year. Finally, 81 qualified volunteers were selected as research objects, including 43 males and 38 females. The BMR, resting lying metabolism rate and resting sitting metabolism rate of the subjects were detected, and the determined BMR was compared with the calculated results: from the adjusted Schofield equation. Results The BMR, resting lying metabolism rate and resting sitting metabolism rate among males were (166.10±22.09), (174.22±24.56), and (179.54±23.35) kJ·m(-2)·h(-1), respectively, which were all higher than those among females were (137.70±20.04), (149.79±19.25), and (167.78±26.02) kJ·m(-2)·h(-1), respectively, ( PBMR of males and females calculated from the adjusted Schofield equation were (160.83±3.93), and (140.29±4.18) kJ·m(-2)·h(-1), respectively, and there was no significantly statistical difference found between the determined BMR and the calculated results from Schofield equation (adjusted) classified by sex, all P values >0.05. Conclusion: The BMR of young adults aged 20-29 years old in Changsha was in the national average level, and the adjusted Schofield equation displayed fine accuracy in predicting BMR of young adults aged 20-29 years old in Changsha.

  19. The coupled code system TORT-TD/ATTICA3D for 3-D transient analysis of pebble-bed HTGR

    International Nuclear Information System (INIS)

    Seubert, A.; Sureda, A.; Lapins, J.; Buck, M.; Laurien, E.; Bader, J.; EnBW Kernkraft GmbH, Philippsburg

    2012-01-01

    This paper describes the time-dependent 3-D discrete-ordinates based coupled code system TORT-TD/ATTICA3D and its application to HTGR of pebble bed type. TORT-TD/ATTICA3D is represented by a single executable and adapts the so-called internal coupling approach. Three-dimensional distributions of temperatures from ATTICA3D and power density from TORT-TD are efficiently exchanged by direct memory access of array elements via interface routines. Applications of TORT-TD/ATTICA3D to three transients based on the PBMR-400 benchmark (total and partial control rod withdrawal and cold helium ingress) and the full power steady state of the HTR-10 are presented. For the partial control rod withdrawal, 3-D effects of local neutron flux redistributions are clearly identified. The results are very promising and demonstrate that the coupled code system TORT-TD/ATTICA3D may represent a key component in a future comprehensive 3-D code system for HTGR of pebble bed type. (orig.)

  20. Nuclear Energy Institute (NEI) summary

    International Nuclear Information System (INIS)

    2001-01-01

    The Nuclear Energy Institute (NEI) provided a brief presentation on the state of energy demand in the United States and discussed the improving economics for new nuclear power plants. He discussed the consolidation of companies under deregulation and the ability of these larger companies to undertake large capital projects such as nuclear power plant construction. He discussed efforts under way to support a new generation of plants but noted that there needs to be greater certainty in the licensing process. He discussed infrastructure challenges in terms of people, hardware, and services to support new and current plants. He stated that there needs to be fair and equitable licensing fees and decommissioning funding assurance for innovative modular designs such as the PBMR. He concluded that NRC challenges will include resolving 10 CFR Part 52 implementation issues, establishing an efficient and predictable process for siting, COL permits and inspection, and an increasing regulatory workload

  1. Research nuclear reactor RA - Annual report 1992; Istrazivacki nuklearni reaktor RA - Izvestaj za 1992. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-12-01

    Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data. [Serbo-Croat] Godisnji izvestaj o radu nuklearnog reaktora RA za 1992. godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA. Prvi deo sadrzi 8 priloga, koji opisuju rad reaktora i poslove sluzbi za odrzavanje opreme i komponenti, finansijski izvestaj, kadrovsku strukturu osoblja reaktora. Drugi prilog (B) je rad Z. Vukadina 'Recurrence formulas for evaluating expansion series of depletion functions' objavljen u casopisu Kerntechnik, 1991. Drugi deo izvestaja o poslovima zastite od zracenja sadrzi 4 priloga sa podacima radijacione kontrole radne sredine i okoline reaktora, opis poslova dekontaminacije i sakupljanja radioaktivnih materija, kao i meteoroloske podatke.

  2. RA research reactor - properties and experimental capabilities; Istrazivacki reaktor RA - Tehnicke karakteristike i eksploatacione mogucnosti

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-05-15

    The brief survey of the Reactor RA exploitation experience, as well as the reactor equipment state, after 18 years of operation is presented. The results of efforts spent on reactor characteristics improvement in order to ensure safe and reliable reactor operation for next 15-20 years, are described. Prikazani su fragmenti iz eksploatacije reaktora kao i stanje opreme, posle 18 godina rada. Na kraju je dat prikaz sta je preduzeto i sta se preduzima da se poboljsaju karakteristike i poveca sigurnost i bezbednost rada za sledecih 15-20 godina.

  3. Reproduction of the RA reactor fuel element fabrication; Reprodukcija izrade gorivnog elementa za reaktor RA

    Energy Technology Data Exchange (ETDEWEB)

    Novakovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This document includes the following nine reports: Final report on task 08/12 - testing the Ra reactor fuel element; design concept for fabrication of RA reactor fuel element; investigation of the microstructure of the Ra reactor fuel element; Final report on task 08/13 producing binary alloys with Al, Mo, Zr, Nb and B additions; fabrication of U-Al alloy; final report on tasks 08/14 and 08/16; final report on task 08/32 diffusion bond between the fuel and the cladding of the Ra reactor fuel element; Final report on task 08/33, fabrication of the RA reactor fuel element cladding; and final report on task 08/36, diffusion of solid state metals. Ovaj rad sadrzi devet priloga: 1. Zavrsni izvestaj o podzadatku 08/12, ispitivanje elementa goriva reaktora RA; 2. Koncepcija izrade gorivnog elementa reaktora RA; 3. Ispitivanje mikrostrukture gorivnog elementa reaktora RA; 4. Zavrsni izvestaj o podzadatku 08/13, dobijanje binarnih legura urana sa legirajucim komponentama Al, Mo, Zr, Nb i B; 5. Dobijanje legure U-Al; 6. Zavrsni izvestaj o podzadacima 08/14 i 08/16; 7. Zavrsni izvestaj o podzadatku 08/32, difuziona veza goriva i kosuljice gorivnog elementa reaktora RA; 8. Zavrsni izvestaj o podzadatku 08/33, izrada kosuljice gorivnog elementa reaktora RA; 9. Zavrsni izvestaj o podzadatku 08/36, difuzija kod metala u cvrstom stanju.

  4. Influence of dissolved product gas on organism retention in biogas tower reactors; Der Einfluss geloester Produktgase auf den Organismenrueckhalt in Biogas-Turmreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Pietsch, T.; Maerkl, H. [Technische Univ. Hamburg-Harburg, Hamburg (Germany). Arbeitsbereich Bioprozess- und Bioverfahrenstechnik

    1999-07-01

    In biogas tower reactors, considerable oversaturations of CO{sub 2} dissolved in molecular form in the liquid phase can occur, compared to the thermodynamic steady state with the gas phase. In buildings of low height, upflow designs cause biological CO{sub 2} production along the reactor to saturate the liquid phase with carbonic acid, and also cause the pH value increasing from acid degradation to bind CO{sub 2} in the form of hydrogen carbonate HCO{sup -}{sub 3}. Where buildings are very high, the liquid phase becomes degassed through a decrease in CO{sub 2} partial pressure because of decreasing hydrostatic pressure along the length of the reactor. Rising gas bubbles in the liquid phase as well as enclosed gas bubbles in biomass particles slow down their sedimentation considerably and can result in flotation of biomass particles owing to gas expansion from declining hydrostatic pressure. A sedimentation characteristics for biomass under decreasing hydrostatic pressure is given. Conditions critical to biomass retention are energy input into CO{sub 2}-oversaturated liquids as well as dynamically rapid drops in pH value owing to associated CO{sub 2} degassing. (orig.) [German] In Biogas-Turmreaktoren koennen erhebliche Uebersaettigungen von molekular geloestem CO{sub 2} in der Fluessigphase gegenueber dem thermodynamischen Gleichgewichtszustand mit der Gasphase auftreten. Bei geringer Bauhoehe fuehrt bei upflow-Konzepten die biologische CO{sub 2}-Produktion entlang des Reaktors zu einer Aufsaettigung der Fluessigphase mit Kohlensaeure und der durch Saeureabbau ansteigende pH-Wert zu einer Bindung des CO{sub 2} in Form des Hydrogencarbonats HCO{sub 3}{sup -}. Sehr grosse Bauhoehen fuehren zu einer Entgasung der Fluessigphase durch Abnahme des CO{sub 2}-Partialdruckes aufgrund des abnehmenden hydrostatischen Druckes entlang der Reaktorhoehe. Aufsteigende Gasblasen in der Fluessigphase sowie eingeschlossene Gasblasen in Biomassepartikeln mindern deren

  5. Study of indicators aggregation techniques for the selection of a new nuclear reactor for Mexico; Estudio de tecnicas de agregacion de indicadores para la seleccion de un nuevo reactor nuclear para Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Barragan M, A.M.; Martin del Campo M, C. [UNAM, Facultad de Ingenieria, 04510 Mexico D.F. (Mexico)]. e-mail: ale_bar_m@yahoo.com.mx

    2007-07-01

    A study on several aggregation techniques that can be used as multi criteria analysis methods, like important part of the methodology developed for the selection of a nuclear reactor for Mexico is described. In an arbitrary way three reactors were selected to be compared, these they are the AP1000 (Advance Passive from 1000 MWe), the PBMR (Pebble Bed Modular Reactor) and the GT-MHR (Gas Turbine Modular Helium). The evaluation approaches were classified in three categories: Economic, Socio-political and of safety and environment. In each category they were defined the more important evaluation indicators and then it was built a matrix with those values of each reactor. The four studied aggregation methods are described: Normalization, Linear deliberation, Fuzzy Logic and AHP (Analytic Hierarchy Process). The well-known aggregation mechanisms are those that are obtained of the lineal deliberation and of the normalization, which have demonstrated to give good results before the simplicity of their use. The fuzzy logic has the advantage that it allows to manage qualitative and quantitative information simultaneously without the aggregation problems that are presented since in a conventional system the semantic pattern on that is based, it is provided by the theory of the diffuse groups that has demonstrated in other areas of the knowledge a better approach to the reality, when admitting that the nature has shades and that the decisions take in function of a wide range of possibilities and of approaches in contradictory occasions or in conflict, all equally worth. The Analytic Hierarchical Process (AHP) that consists in formalizing the intuitive understanding of a multi criteria complex problem, by means of the construction of a hierarchical model that allows the decision agent to structure the problem in visual form, giving him the form of a hierarchy of attributes (global objective of the problem, approaches and alternative). Finally, using the matrix of initiators

  6. ILK statement on the consequences of the Chernobyl accident. Taking stock after twenty years

    International Nuclear Information System (INIS)

    2006-01-01

    The Chernobyl reactor accident was the consequence of a reactor design which was not inherently safe, and of a lack of 'safety culture'. The RBMK-type reactor (a Russian graphite-moderated light water reactor design: reaktor bolshoi moshnosty kanalny=high-power channel reactor) had not been designed to a satisfactory safety level, and the operating staff were not informed on the weak spots in plant design. The combination of these factors caused the worst nuclear accident, completely destroying the reactor. The consequences may be seen as the product of two severe accidents superimposed upon each other: the explosion of the reactor, and core melt-down associated with an intense, persistent fire of the graphite moderator. The Statement contains analyses of these points: Release, Propagation and Deposition of Radioactive Materials; Protective Measures; Impact on the Environment and Agriculture; Assessment of Radiation Exposure; Health Impact; Psychological and Societal Impacts; Potential Residual Risks. (orig.)

  7. Anaerobic treatment of sulfate-containing wastewater from distilleries

    International Nuclear Information System (INIS)

    Stadlbauer, E.A.; Oey, L.N.; Weber, B.

    1994-01-01

    Bioprocess evaluation of a staged arrangement of a Pulse Driven Loop Reaktor (PDLR) and a Pulsed Anaerobic Filter (PAF) using highly polluted cherry slops as industrial wastewater shows a COD removal efficiency of 80-90% at loading rates of 8-4 kg COD/(M 3 .d). Contamination of cherry slops by sulfate (2 g/l) and copper (150-200 mg/l) reduces COD degradation to 40-50 percent. A pulsed anaerobic baffled reactor was envisaged as a corrective tool to improve mineralisation in the presence of sulfate-rich substrates by confining sulfate reducing bacteria to the first 4 chambers of the reactor. Phasing slightly improves COD degradation yield, but is not sufficient for stable process performance. Consequently, the use of lactic acid in stead of sulfuric acid in cherry-fermentation was suggested as a preventive method to avoid sulphide-induced digester failure. (orig.) [de

  8. Neutronic characterization of the SAFARI-1 material testing reactor - HTR2008-58155

    International Nuclear Information System (INIS)

    Makgopa, B. M.; Belal, M.; Strydom, W. J.

    2008-01-01

    This work presents a neutronic analysis of the core in the South African Fundamental Atomic Research Installation (SAFARI-1) for future Pebble Bed Modular Reactor (PBMR) fuel irradiation experiments. Monte Carlo simulation of the core with and without the rig has been performed. The results show a negligibly small reactivity worth of the rig, which is expected, due to the small amount of heavy metal loading in the pebble and the low fuel enrichment. This effect will be further investigated when the rig is extended to include more than one fuel pebble. Results further show perturbations in the neutron and photon flux as well as the power distribution in core position B6. A 50% thermal neutron flux depression is observed in position B6 due to the insertion of the rig. A 60% increase in axial photon heating values is also observed in position B6. The neutron and photon flux and power distributions in the other in-core irradiation positions (D6 and F6) are slightly affected by the insertion of this rig. Fluxes and power distributions in positions D6 and F6 will be studied in detail when they are loaded with isotope production rigs. (authors)

  9. HTR core physics analysis at NRG

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

    2002-01-01

    Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

  10. Distance determination of NPP and oil reservoir on enhanced oil recovery based on heat loss and safety in view point

    International Nuclear Information System (INIS)

    Erlan Dewita; Dedy Priambodo; Sudi Ariyanto

    2013-01-01

    EOR is a method used to increasing oil recovery by injecting material or other to the reservoir. There are 3 EOR technique have been used in the world, namely thermal injection, chemical injection dan Miscible. Thermal injection method is the method most widely used in the world, however, one drawback is the loss of heat during steam distribution to the injection wells. In Indonesia, EOR application has been successfully done in the field of Duri, Chevron uses steam injection method, but still use petroleum as a fuel for steam production. In order to save oil reserves, it was done the introduction of co-generation nuclear power plants to supply some of the heat of nuclear power plants for EOR processes. In cogeneration nuclear power plant, the safety aspect is main priority. The purpose of the study was to evaluate the distance NPP with oil wells by considering heat loss and safety aspects. The method of study and calculations done using Tempo Cycle program. The study results showed that in the distance of 400 meter as exclusion zone of PBMR reactor, with pipe insulation thickness 1 in, the amount of heat loss of 277, 883 kw, while in pipe isolation thickness 2 in, amount of heat loss became 162,634 kw and with isolation thickness 3 in, amount of heat loss 120,767 kw., heat loss can be overcome and provide insulation pipes and improve the quality of saturated steam into superheated. (author)

  11. Pebble Bed Modular Reactor. Executive summary, Issue D

    International Nuclear Information System (INIS)

    1979-01-01

    The PBMR Project has been under active investigation by Eskom, as part of the Integrated Electricity Planning process, since 1993. The overall objectives of these investigations have been to establish whether such a system could form part of Eskom's expansion planning and what specific advantages it would bring over current options. This would include a technical performance and economic evaluation associated with the project. A comprehensive evaluation was performed as to the international interest existing within this field of technology, including the availability of this technology. The first phase of this investigation is now complete and the results have been compiled in a comprehensive set of technical and costing reports. A basic engineering simulator is being developed which provides engineering design support. The present results of this study show that the design has been established in enough detail to support key safety studies, confirm operating limits and estimated costing. The costing includes: - Production costs for a single module and a unit (unit = 10 x module) cost; - Operation and maintenance costing; - Fuel plant costing ? Design and development costing The studies showed that the technology required for the design has been demonstrated adequately to avoid fundamental technical risk. The increased level of inherent safety (over current designs) is fundamental to the cost reductions achieved over other nuclear designs. By demonstrating a catastrophe free design the requirements for both safety grade backup systems and an off-site emergency plan are removed

  12. Parní generátor reaktoru ESFR

    OpenAIRE

    Bátěk, David

    2012-01-01

    Tato diplomová práce se zabývá parním generátorem pro reaktory ESFR (Evropský Rychlý Sodíkem Chlazený Reaktor), jenž je vyhříván tekutým sodíkem. V úvodních kapitolách jsou teoretické informace o parametrech ESFR a porovnání s výměníky tepla v jaderných elektrárnách pracujících na stejném principu (sodík jako chladivo). Následuje návrhová část, kam patří úvod do problematiky výpočtu, volba materiálu a koncepce výměníku a samotná výpočtová část, která zahrnuje tepelný, hydraulický a pevnostní ...

  13. Structural analysis of aircraft impact on a nuclear powered ship

    International Nuclear Information System (INIS)

    Dietrich, R.

    1976-01-01

    As part of a general safety analysis, the reliability against structural damage due to an aircraft crash on a nuclear powered ship is evaluated. This structural analysis is an aid in safety design. It is assumed that a Phantom military jet-fighter hits a nuclear powered ship. The total reaction force due to such an aircraft impact on a rigid barrier is specified in the guidelines of the Reaktor-Sicherheitskommission (German Safety Advisory Committee) for pressurized water reactors. This paper investigates the aircraft impact on the collision barrier at the side of the ship. The aircraft impact on top of the reactor hatchway is investigated by another analysis. It appears that the most unfavorable angle of impact is always normal to the surface of the collision barrier. Consequently, only normal impact will be considered here. For the specific case of an aircraft striking a nuclear powered ship, the following two effects are considered: Local penetration and dynamic response of the structure. (Auth.)

  14. PENGELOLAAN LIMBAH PETERNAKAN SAPI UNTUK MENINGKATKAN KAPASITAS PRODUKSI PADA KELOMPOK TERNAK PATRA SUTERA

    Directory of Open Access Journals (Sweden)

    Danang Dwi Saputro

    2014-02-01

    Full Text Available Kelompok ternak Patra Sutera di Desa Ledok Kecamatan Sambong Kabupaten Blora yang berdiri pada tahun 2013 telah mempunyai sapi 8 ekor yang berada di kandang komunal yang dikelola oleh 6 anggota kelompok. Dalam satu hari setiap ekor sapi dapat menghasilkan limbah padat sebanyak 20-30 kg dan limbah cair sebanyak 100-150 liter yang selama ini belum dikelola dengan baik. Limbah dari kegiatan ternak belum terolah dengan baik dan dibuang ke lingkungan sehingga menimbulkan dampak negatif bagi kesehatan masyarakat sekitar kandang. Salah satu cara untuk mengatasi kondisi ini adalah dengan memberikan pelatihan keterampilan atau pendampingan bagaimanakah teknik pembuatan pupuk organic dan reaktor biogas sederhana, mengoperasikan, serta memanfaatkan gas yang dihasilkan. Dalam kegiatan ini akan diberikan pelatihan keterampilan bagaimana cara mengolah limbah ternah untuk dijadikan pupuk dan pestisida organik,serta pengelolaan biodigester. Dari kegiatan ini Anggota kelompok ternak Patra Sutera mendapat pengetahuan dan mengolah limbah kotoran ternak (padat dan cair yang keliar dari biodigester menjadi pupuk yang lebih bermanfaat.

  15. Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

    International Nuclear Information System (INIS)

    Irianto, Djoko; Kanji, T.; Kukita, Y.

    1996-01-01

    An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

  16. Slovak brown coals as a feedstock for the active coke production

    Directory of Open Access Journals (Sweden)

    Sobolewski Aleksander

    1998-09-01

    Full Text Available V èlánku sa venuje pozornos možnostiam výroby aktívneho koksu zo slovenského hnedého uhlia Bane Cíge¾. Príprava aktívneho koksu bola uskutoènená v laboratórnych podmienkach a v poloprevádzke. Surové uhlie sa podrobilo vysokoteplotnej pyrolýze v retorte s pevným roštom, v klasickej rotaènej peci, ako aj vo fluidnom reaktore. Následne boli urèované adsorpèné charakteristiky získaného aktívneho koksu, ktoré sa porovnávali s charakteristikami komerène vyrábaného koksu. V príspevku sa taktiež diskutujú možnosti aplikácie pripraveného aktívneho koksu v technológiách ochrany životného prostredia.

  17. Independent CO{sub 2} loop for cooling the samples irradiated in vertical experimental channels of the RA reactor, Vol. I; Nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA, Album I

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M; Pavlovic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-06-15

    Project 'independent CO{sub 2} loop for cooling the samples irradiated in the vertical experimental channels of the RA reactor' is presented in two volumes: volume I - head of the low temperature coolant loop for reactor RA, and volume II - Outer low-temperature reactor coolant loop. Volume I includes: the design specifications for the head of the low-temperature coolant loop, technical description, thermal calculation, calculations of mechanical loads, antireactivity and activation of the components of the coolant loop head, engineering schemes and drawings, cost estimation data. [Serbo-Croat] Projekat 'Nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u vertikalnim eksperimentalnim kanalima reaktora RA', sastoji se od dva albuma: album I - Glava niskotemperaturno rashladne petlje za reaktor RA, album II - Spoljno kolo niskotemperaturne rashladne petlje za reaktora. Album I sadrzi projektni zadatak glave niskotemperaturne petlje, tehnicki opis, termicki proracun, proracun mehanickih naprezanja, antireaktivnosti i aktivacije kontrukcionih elemenata glave petlje, konstrukcione seme i crteze glave petlje, predracun.

  18. Outlook on to fuel cycle perspectives at WWER-440

    International Nuclear Information System (INIS)

    Stech, S.; Bajgl, J.

    2005-01-01

    Current internal fuel cycle in NPP Dukovany 4x440 MWe is shortly characterized with new types of fuel assemblies and advanced fuel cycles which have been introduced in the last years. The modernization activities accomplished until now might be extrapolated to the further period in fuel design - mechanic, thermal-hydraulic and neutronic respectively - with additional increase in fuel enrichments and burnups on the way to the 6-year cycle. Reaktor power up rating together with Unit thermal efficiency improvements could bring an increase in the electric output to the value nearly 500 MWe. The reasons are given for long-term cooperation with Fuel Supplier and Plant Designer in the area of fuel cycle as well as in Unit Design Basis. All innovations mentioned in the article including future fuel and fuel cycle changes might be a quite realistic perspective at the end of the first decade of the new century (Authors)

  19. Design of fuel loading for Bohunice V-1 Unit 2 reaktor for fuel cycle No.19

    International Nuclear Information System (INIS)

    Majercik, J.

    1998-01-01

    The report contains description of the design of fuel loading for the fuel cycle No. 19 in the V-1 Bohunice Unit 2 reactor. Input data and computer codes used for the development of the design are shown. The fuel loading is characterized by the assortment of the fuel loaded and by the scheme of re shuffling of assemblies in the core. An evaluation of basic neutronic core parameters as relates to the compliance with safety criteria is a part of the report as well

  20. RB Research nuclear reactor, 30 years of operation; Istrazivacki nuclearni reaktor RB, povodom 30 godina rada

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Stefanovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1988-06-15

    Paper describes utilization, modifications and changes of construction and control-safety systems done at the RB reactor during 30 years of operation. Experiments performed at the reactor are summarized, new reactor equipment is described and the future plans are shown. Rad prikazuje eksploataciju reaktora RB tokom 30 godina rada, modifikacije i izmene u konstruktivnim i upravljacko-sigurnosnim sistemima. Sumirani su eksperimenti izvedeni na njemu, prikazana je nova oprema i planovi za buduci rad.

  1. Nuclear RB research reactor. Thirty years of anniversary; Istrazhivacki nuklearni reaktor RB. Povodom 30 godina rada

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Stefanovic, D [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

    1988-07-01

    Nuclear research reactor RB in the Nuclear Engineering Laboratory - NET at the 'Boris Kidric' Institute of Nuclear Sciences in Vinca is the first reactor system built in Yugoslavia in 1958. This year is the thirtieth anniversary of the RB reactor operation, which has survived a series of modifications trying to follow a contemporary nuclear research directions. This report describes its basic technical characteristics and experimental possibilities. Especially, the modifications in the last 25 years are underlined, the experiences gained, and new plans for the future are presented. (author)

  2. German safety engineering in comparison to the Chernobyl reactor; Deutsche Sicherheitstechnik im Vergleich zum Tschernobyl-Reaktor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-04-15

    In Germany the Russian RBMK would not have been licensed, since the safety standards in Germany were superior to those in the former USSR. The most significant differences - the containment, the control rod system, the void coefficient and the emergency cooling system are shortly summarized. The consequences for the population are cumulative environmental radiation exposure are reported using the official data of IAEA, WHO and UNDP.

  3. Preliminary Core Design Analysis of a 200MWth Pebble Bed-type VHTR

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Noh, Jae Man

    2007-01-01

    This paper intends to suggest the preliminary core design analysis of a VHTR for a hydrogen production. The nuclear hydrogen system that utilizes the high temperature heat generated from the VHTR is a promising candidate for a cost effective, safe and clean supply of hydrogen in the age of hydrogen economy. Among two candidate VHTR cores, that is, a prismatic modular reactor (PMR) and a pebble bed-type reactor (PBR), we focus on the design of a 200MWth PBR (hereinafter PBR200) in this paper. Here, the 200MWth power is selected for a demonstration plant. The core configuration of the PBR200 is similar to the PBMR (Pebble Bed Modular Reactor, 400MWth) of South Africa, but the overall dimension of the reactor system is scaled-down. This paper is to suggest two candidate PBR200 cores. One is an annular core with an inner reflector (PBR200-CD1) which was presented at IWRES07, and the other is a cylindrical core without an inner reflector (PBR200-CD2)

  4. Establishment of a South African nuclear science exhibition centre

    Energy Technology Data Exchange (ETDEWEB)

    Lekwe, K.G.; Stander, G.; Faanhof, A. [South African Nuclear Energy Cooperation, P O Box 582, Pretoria (South Africa)

    2008-07-01

    After an initial survey undertaken by the South African Nuclear Energy Corporation (Necsa), one of the findings was that nuclear knowledge is virtually non-existent amongst the general public, including school children, throughout the country. The Department of Education (DoE) is currently in the process of introducing Nuclear as part of the school curriculum, which would require a collective effort between the schools and all the Nuclear Institutions in the country. Necsa as well as other nuclear industries have the responsibility to promote public awareness, appreciation and understanding of science and nuclear science in particular. Necsa is leading the national initiative to establish the nuclear science centre which would amongst others guide a person from the very basics of nuclear science to present and future applications thereof. The nuclear science centre will include information on the SAFARI-1 reactor, the Koeberg power reactor, the Pebble Bed Modular Reactor (PBMR), particle accelerators, food preservation, medical applications, etc. This paper will give the overview of the centre as well as its objectives thereof. (authors)

  5. Use of active and passive microwave remote sensing for soil moisture estimation through corn

    International Nuclear Information System (INIS)

    O'Neill, P.E.; Chauhan, N.S.; Jackson, T.J.

    1996-01-01

    Over the past several years NASA, USDA, and Princeton University have collaborated to conduct hydrology field experiments in instrumented research watersheds in Pennsylvania and Oklahoma with a goal of characterizing the spatial and temporal variability of soil moisture using microwave sensors. As part of these experiments, L-band radar data from both truck and aircraft sensors were used to validate the performance of a vegetation scattering model in which discrete scatter random media techniques were employed to calculate vegetation transmissivity and scattering. These parameters were then used in a soil moisture prediction algorithm based on a radiative transfer approach utilizing aircraft passive microwave data from the L-band PBMR and ESTAR radiometers. Soil moisture was predicted in both experiments for several large corn fields which represented the densest vegetation canopies of all the test fields. Over the 20 per cent change in soil moisture encountered in the experiments, the match of predicted to measured soil moisture was excellent, with an average absolute error of about 0 · 02 cm 3 cm −3 . (author)

  6. Establishment of a South African nuclear science exhibition centre

    International Nuclear Information System (INIS)

    Lekwe, K.G.; Stander, G.; Faanhof, A.

    2008-01-01

    After an initial survey undertaken by the South African Nuclear Energy Corporation (Necsa), one of the findings was that nuclear knowledge is virtually non-existent amongst the general public, including school children, throughout the country. The Department of Education (DoE) is currently in the process of introducing Nuclear as part of the school curriculum, which would require a collective effort between the schools and all the Nuclear Institutions in the country. Necsa as well as other nuclear industries have the responsibility to promote public awareness, appreciation and understanding of science and nuclear science in particular. Necsa is leading the national initiative to establish the nuclear science centre which would amongst others guide a person from the very basics of nuclear science to present and future applications thereof. The nuclear science centre will include information on the SAFARI-1 reactor, the Koeberg power reactor, the Pebble Bed Modular Reactor (PBMR), particle accelerators, food preservation, medical applications, etc. This paper will give the overview of the centre as well as its objectives thereof. (authors)

  7. Efforts of development on the next generation nuclear reactor in the Mitsubishi Heavy Industries, Ltd

    International Nuclear Information System (INIS)

    Mukai, Hiroshi

    2002-01-01

    At present, the Mitsubishi Heavy Industry, Ltd. (MHI) enters to development on APWR+ for a large-scale reactor, AP1000 and pebble bed modular reactor (PBMR) for middle- and small-scale one, and innovative one, under cooperation of power industries, manufacturers and institutes in and out of Japan. On APWR+, MHI occupies the most advanced position of conventional large-scale route, intends to carry out further upgrading of large capacity on a base of already developed 1500 MWe class APWR, and aims at further upgrading of economical efficiency. On the other reactor, as it becomes possible to perform value addition specific to the small-scale reactor with smaller output, it is planned to overcome its scale demerit by introducing more innovative techniques. And, on AP1000, it is intended to remove dynamic safety system by introducing a static one, to upgrade simplification of apparatus and reliability of safety system and to reduce its human factors. In addition, here was described on the next generation nuclear reactors under development. (G.K.)

  8. Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

    International Nuclear Information System (INIS)

    Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

    2014-01-01

    This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

  9. PENGARUH NILAI BAKAR TERHADAP INTEGRITAS KELONGSONG ELEMEN BAKAR TRIGA 2000

    Directory of Open Access Journals (Sweden)

    K.A. Sudjatmi

    2015-04-01

    Full Text Available Bentuk elemen bakar reaktor TRIGA Bandung adalah silinder padat yang merupakan campuran homogen paduan uranium dan zirkonium hidrida. Pada saat reaktor beroperasi, suhu elemen bakar akan bertambah, akibatnya akan menaikan tekanan gas-gas yang ada di dalam kelongsong elemen bakar. Tekanan gas yang timbul dalam kelongsong elemen bakar merupakan penjumlahan tiga komponen tekanan yaitu tekanan akibat udara yang terperangkap antara kelongsong dengan bahan bakar, tekanan oleh gas hasil fisi yang terbentuk dari elemen bakar dan tekanan yang berasal dari pemisahan hidrogen dari paduan zirkonium hidrida. Gas hasil fisi yang terbentuk oleh bahan bakar sebanding dengan besarnya fraksi bakar oleh setiap elemen bakar dalam teras reaktor. Semakin besar fraksi bakar elemen bakar, semakin besar gas gas hasil fisi yang dihasilkannya, akibatnya semakin besar tekanan di dalam kelongsong yang disebabkan oleh gas gas hasil fisi tersebut. Perhitungan jumlah gas-gas hasil fisi dalam kelongsong yang merupakan fungsi dari nilai bakar dilakukan dengan menggunakan program ORIGEN-2. Program ORIGEN-2 adalah kode komputer yang banyak digunakan untuk menghitung hasil fisi, peluruhan dan pengolahan bahan radioaktif. Tampang lintang, presentase timbulnya hasil fisi, data peluruhan, dan data lainnya yang diperlukan disediakan dalam pustaka data selama eksekusi program. Dari hasil perhitungan dapat disimpulkan bahwa tekanan gas yang diakibatkan oleh gas hasil fisi adalah 4,13 10-3 psi dan tekanan gas yang diakibatkan udara yang terjebak di dalam kelongsong adalah 56,6 psi, yang mengakibatkan tegangan pada kelongsong sebesar 2080 psi dan nilai ini jauh lebih kecil dari setengah tegangan luluh bahan kelongsong sebesar 12.000 psi pada temperatur 750 oC atau sekitar 40.000 psi pada temperatur 138 oC. Akhirnya dapat disimpulkan bahwa dilihat dari sisi nilai bakar, maka elemen bakar layak digunakan sampai mencapai nilai bakar maksimum. Kata kunci : TRIGA, nilai bakar, elemen bakar

  10. ANALISIS EFEK KECELAKAAN WATER INGRESS TERHADAP REAKTIVITAS DOPPLER TERAS RGTT200K

    Directory of Open Access Journals (Sweden)

    Zuhair Zuhair

    2015-03-01

    Full Text Available Dalam high temperature reactor, koefisien reaktivitas temperatur yang didesain negatif menjamin reaksi fisi dalam teras tetap berada di bawah kendali dan panas peluruhan tidak akan pernah melelehkan bahan bakar yang menyebabkan terlepasnya zat radioaktif ke lingkungan. Namun masuknya air (water ingress ke dalam teras reaktor akibat pecahnya tabung penukar panas generator uap, yang dikenal sebagai salah satu kecelakaan dasar desain, dapat mengintroduksi reaktivitas positif dengan potensi bahaya lainnya seperti korosi grafit dan kerusakan material struktur reflektor. Makalah ini akan menganalisis efek kecelakaan water ingress terhadap reaktivitas Doppler teras RGTT200K. Kapabilitas koefisien reaktivitas Doppler untuk mengkompensasi reaktivitas positif yang timbul selama kecelakaan water ingress akan diuji melalui serangkaian perhitungan dengan program MCNPX dan pustaka ENDF/B-VII untuk perubahan temperatur bahan bakar dari 800K hingga 1800K. Tiga opsi kernel bahan bakar UO2, ThO2/UO2 dan PuO2 dengan tiga model kisi bahan bakar pebble di teras reaktor diterapkan untuk kondisi water ingress dengan densitas air dari 0 hingga 1.000 kg/m3. Hasil perhitungan memperlihatkan koefisien reaktivitas Doppler tetap negatif untuk seluruh opsi bahan bakar yang dipertimbangkan bahkan untuk posibilitas water ingress yang besar. Efek water ingress lebih kuat pada model kisi dengan fraksi packing lebih rendah karena lebih banyak volume yang tersedia untuk air yang memasuki teras reaktor. Efek water ingress juga lebih kuat di teras uranium dibandingkan teras thorium dan plutonium sebagai konsekuensi dari fenomena Doppler dimana absorpsi neutron di daerah resonansi 238U lebih besar daripada 232Th dan 240Pu. Secara keseluruhan dapat disimpulkan bahwa, koefisien Doppler teras RGTT200K mampu mengkompensasi insersi reaktivitas yang diintroduksi oleh kecelakaan water ingress. Teras RGTT200K dengan bahan bakar UO2, ThO2/UO2 dan PuO2 dapat mempertahankan fitur keselamatan

  11. Is the rate of metabolic ageing and survival determined by Basal metabolic rate in the zebra finch?

    Science.gov (United States)

    Rønning, Bernt; Moe, Børge; Berntsen, Henrik H; Noreen, Elin; Bech, Claus

    2014-01-01

    The relationship between energy metabolism and ageing is of great interest because aerobic metabolism is the primary source of reactive oxygen species which is believed to be of major importance in the ageing process. We conducted a longitudinal study on captive zebra finches where we tested the effect of age on basal metabolic rate (BMR), as well as the effect of BMR on the rate of metabolic ageing (decline in BMR with age) and survival. Basal metabolic rate declined with age in both sexes after controlling for the effect of body mass, indicating a loss of functionality with age. This loss of functionality could be due to accumulated oxidative damage, believed to increase with increasing metabolic rate, c.f. the free radical theory of ageing. If so, we would expect the rate of metabolic ageing to increase and survival to decrease with increasing BMR. However, we found no effect of BMR on the rate of metabolic ageing. Furthermore, survival was not affected by BMR in the males. In female zebra finches there was a tendency for survival to decrease with increasing BMR, but the effect did not reach significance (PBMR on the rate of functional deterioration with age, if any, was not strong enough to influence neither the rate of metabolic ageing nor survival in the zebra finches.

  12. STUDI EKSPERIMEN PEMILIHAN BIOMASSA UNTUK MEMPRODUKSI GAS ASAP CAIR ( LIQUID SMOKE GASES SEBAGAI BAHAN PENGAWET

    Directory of Open Access Journals (Sweden)

    Sugeng Slamet

    2015-04-01

    Full Text Available ABSTRAK Pengertian umum asap cair merupakan suatu hasil destilasi atau pengembunan dari uap hasil pembakaran tidak langsung maupun langsung dari bahan yang banyak mengandung karbon dan senyawa- senyawa lain. Bahan baku yang banyak digunakan untuk membuat asap cair adalah kayu, bongkol kelapa sawit, ampas hasil penggergajian kayu, dan lain-lain. Pembuatan asap cair menggunakan metode pirolisis yaitu peruraian dengan bantuan panas tanpa adanya oksigen atau dengan jumlah oksigen yang terbatas. Biasanya terdapat tiga produk dalam proses pirolisis yakni: gas, pyrolisis oil, dan arang, yang mana proporsinya tergantung dari metode pirolisis, karakteristik biomassa dan parameter reaksi. Metode yang dilakukan diawali dengan melakukan rancang bangun unit pirolisator lengkap dengan perangkat kondensor dengan pipa tembaga tipe spiral untuk memproduksi gas asap cair dari bahan biomassa yang dipilih yaitu tempurung kelapa dan sampah organik. Metode Pirolisis yang merupakan proses reaksi penguraian senyawa-senyawa penyusun kayu keras menjadi beberapa senyawa organik melalui reaksi pembakaran kering pembakaran tanpa oksigen. Reaksi ini berlangsung pada reaktor pirolisator dengan variasi temperatur 150oC, 250oC dan 300oC selama 8 jam pembakaran. Asap hasil pembakaran dikondensasi dengan kondensor yang berupa pipa tembaga melingkar. Hasil dari proses pirolisis diperoleh tiga produk yaitu asap cair, tar, dan arang. Kondensasi dilakukan dengan pipa atau koil melingkar yang dipasang dalam bak pendingin. Air pendingin dapat berasal dari air hujan yang ditampung dalam bak penampungan. Hasil yang diperoleh dari penelitian ini adalah biomassa tempurung kelapa menghasilkan jumlah senyawa fenol lebih besar 30-33%. Hal ini menunjukkan bahwa pada jenis biomassa ini lebih unggul dalam fungsi sebagai antioksidan, karena kaya akan kandungan senyawa fenol, sehingga lebih optimal dalam hal menghambat kerusakan pangan dengan cara mendonorkan hidrogen. Sedangkan biomassa cangkang

  13. An endothermic chemical process facility coupled to a high temperature reactor. Part I: Proposed accident scenarios within the chemical plant

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

    2012-01-01

    Highlights: ► The paper identifies possible transient and accident scenarios in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Key accidents scenarios were investigated through qualitative reasoning. ► The accidents were found to constitute loss of heat sink event for the nuclear reactor. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Quantitative study of the possible operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven unique case studies are proposed based on a thorough review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. The qualitative parameters of each case study are outlined as well as the basis in literature. A previously published modeling scheme is described and adapted for application as a simulation platform for these transient events. The results of the quantitative case studies are described within part II of this paper.

  14. Základy fúzní energetiky II - Základní fyzika fúzních reaktorů

    Czech Academy of Sciences Publication Activity Database

    Entler, Slavomír; Mlynář, Jan; Dostál, V.

    Srpen (2016), č. článku 14538. ISSN 1801-4399 Institutional support: RVO:61389021 Keywords : Nuclear fusion * fusion reactors * plasma Subject RIV: JF - Nuclear Energetics http://energetika.tzb-info.cz/elektroenergetika/14538-zaklady-fuzni-energetiky-ii-zakladni-fyzika-fuznich-reaktoru

  15. Neutron Diffractometer; Neutronski difraktometar

    Energy Technology Data Exchange (ETDEWEB)

    Zivadinovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    RA nuclear reactor is considered as a relatively strong neutron source producing the thermal neutron flux of about 3x10{sup 13} n/cm{sup 2} sec when operating at nominal power of 6.5 MW. Neutron diffraction method is applied in the field of solid state physics, material science, crystallography, magnetism, nuclear physic. Neutron diffractometer at the RA reactor consists of: system for obtaining collimated neutron beam from the horizontal experimental channel neutron monochromator; goniometer and electronic equipment for measurements and collecting the the measurement data. Nuklearni reaktor RA koji pri radu na snazi od 6,5 MW ima fluks termalnih neutrona oko 3x10{sup 13} n/cm{sup 2} sec predstavlja relativno jak izvor neutrona. Tehnika difrakcije neutrona primenjuje se u istrazivanjima fizike crvstog stanja, strukture materijala, kristalografije, magnetizma, nuklearne fizike. Neutronski difraktometar na reaktoru RA sastoji se od sistema za dobijanje kolimisanog snopa neutrona kroz horizontalni kanal reaktora; neutroskog monohromatora; goniometra i elektronskih uredjaja za merenja i registrovanje rezultata. Ovaj izvestaj sadrzi detaljan opis i seme neutronskog difraktometra sa pratecom opremom i elektronskim komponentama.

  16. Remerschen nuclear power station with BBR pressurized water reactor

    International Nuclear Information System (INIS)

    Hoffmann, J.P.

    1975-01-01

    On the basis of many decades of successful cooperation in the electricity supply sector with the German RWE utility, the Grand Duchy of Luxemburg and RWE jointly founded Societe Luxembourgeoise d'Energie Nucleaire S.A. (SENU) in 1974 in which each of the partners holds a fifty percent interest. SENU is responsible for planning, building and operating this nuclear power station. Following an international invitation for bids on the delivery and turnkey construction of a nuclear power station, the consortium of the German companies of Brown, Boveri and Cie. AG (BBC), Babcock - Brown Boveri Reaktor GmbH (BBR) and Hochtief AG (HT) received a letter of intent for the purchase of a 1,300 MW nuclear power station equipped with a pressurized water reactor. The 1,300 MW station of Remerschen will be largely identical with the Muelheim-Kaerlich plant under construction by the same consortium near Coblence on the River Rhine since early 1975. According to present scheduling, the Remerschen nuclear power station could start operation in 1981. (orig.) [de

  17. Data base formation for important components of reactor TRIGA MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, R; Mavko, B; Kozuh, M [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [Slovenian] V referatu smo prikazali raziskavo, v okviru katere smo za raziskovalni reaktor TRIGA MARK II v Podgorici izoblikovali specificno bazo podatkov. Zbrali smo podatke obratovanja reaktorja od leta 1985 do 1990. Rezultate raziskave dogodkov smo razdelili v dve glavni skupini. V prvo spadajo obratovalni podatki o komponentah, v drugo skupino pa spadajo zagoni oz. zaustavitve reaktorja. Podatke smo ovrednotili z modelom v casovnem prostoru in z modelom na zahtevo. Parametre modelov smo dolocili s klasicno metodo. Opisane baze podatkov so uporabne povsod, kjer so lahko posledice nezanesljivega delovanja sistemov velike. [author].

  18. SAFIRA project B.3.3: in-situ-treatment of contaminated ground water by catalytic oxidation. Final report; Sanierungsforschung in regional kontaminierten Aquiferen (SAFIRA). Projekt B.3.3: In situ-Behandlung von kontaminierten Grundwaessern durch katalytische Oxidation. Teilvorhaben 1: Untersuchungen im Labormassstab. Teilvorhaben 2: Tests in der bench-scale-Anlage und Teilvorhaben 3: Die Erprobung in der Pilotanlage am Modellstandort. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, J.; Haentzschel, D.; Freier, U.; Wecks, M.

    2003-06-27

    A new technology for treatment of contaminated ground water was developed. In this process heterogeneous catalysts (full metal catalyst, mixed oxide catalyst or iron-containing zeolites) in combination with hydrogen peroxide are used. In the reactor catalytic oxidation and aerob biological degradation occur simultaneously. A complete degradation of chlorobenzene was observed in a bench-scale-equipment (2 liter) and also in the pilot plant at the model site located in Bitterfeld (30 liter reactor). The technology can be applied to the ground and waste water treatment. (orig.) [German] Fuer die Behandlung von Grundwaessern, die mit organischen Schadstoffen belastet sind, wurde ein neuartiges Verfahren entwickelt. Bei der katalytischen Oxidation werden heterogene Katalysatoren in Form von Vollmetall-, Mischoxid- und Traegerkatalysatoren in Verbindung mit Wasserstoffperoxid als Oxidationsmittel eingesetzt. In den Katalysereaktoren laufen die heterogen-katalytische Oxidation und der aerob-biologische Abbau nebeneinander ab. Es werden synergistische Effekte erzielt. Mit dem Verfahren wurde in einer bench-scale-Angle (2 Liter) und in der Pilotanlage am Modellstandort in Bitterfeld (30 l Reaktor) der Schadstoff Chlorbenzol vollstaendig umgesetzt. Das Verfahren kann zur Grund- und Abwasserbehandlung eingesetzt werden. (orig.)

  19. Sintesis Furfural dari Bagas Tebu Via Reaksi Hidrolisa dengan Menggunakan Katalis Asam Asetat pada Kondisi Atmosferik

    Directory of Open Access Journals (Sweden)

    Nine Tria Rossa

    2015-12-01

    Full Text Available Telah dilakukan penelitian tentang hidrolisa bagas tebu menggunakan asam asetat sebagai katalis. Sebanyak 30 gram dihidrolisa dalam 300ml air yang mengandung katalis asam menggunakan  asetat sebesar 7 sampai 9% v/v dengan variabel waktu dan temperatur hidrolisa selama 1 sampai 4 jam dan 80oC sampai 103oC pada kondisi atmosferik menggunakan reaktor tipe batch. Penelitian ini bertujuan untuk mengetahui pengaruh temperatur dan waktu hidrolisa, serta konsentrasi katalis asam asetat terhadap perolehan furfural. Kemudian untuk menemukan kondisi paling efisien untuk memproduksi furfural menggunakan Response Surface Methodology (RSM dengan Software Design Expert 7.0.0. Hasil penelitian menunjukkan bahwa dengan penambahan waktu dan temperatur hidrolisa akan meningkatkan perolehan furfural. Perolehan asam asetat juga meningkat rata-rata hingga 2 kali dari konsentrasi asam asetat awal. Hal ini terjadi karena pemutusan gugus acetyl dari fraksi hemiselulosa pada bagas tebu. Perolehan kondisi optimum yakni pada waktu dan temperatur hidrolisa 2 jam dan 103oC, konsentrasi katalis 9%, dengan konsentrasi furfural 4,10 mg/ml dan konsentrasi asam asetat 2,62 mmol/ml.

  20. Pirolisis Campuran Sampah Plastik Polistirena Dengan Sampah Plastik Berlapisan Aluminium Foil (Multilayer

    Directory of Open Access Journals (Sweden)

    Yebi Yuriandala

    2016-04-01

    Full Text Available Sampah plastik yang dulunya merupakan masalah lingkungan, saat ini dapat diubah menjadi bahan bakar alternatif dengan menggunakan proses daur ulang yang memanfaatkan energi panas yaitu pirolisis. Analisis yang digunakan dalam penelitian ini adalah analisis terhadap minyak (liquid yang dihasilkan dari pirolisis sampah plastik Polistitren (PS, plastik berlapisan aluminium foil (kemasan/ multilayer (AL dan campuran plastik tersebut. Penelitian ini dilakukan untuk mendapatkan kuantitas produk dan senyawa kimia yang dihasilkan dari pirolisis sampah plastic PS, kemasan dan campuran keduanya. Penelitian dilakukan dengan menempatkan 50 gram PS (PS, 50 gram plastik berlapisan aluminium foil (AL (multi layer,dan PS dengan campuran 10%, 20%, 30%, 40% AL didalam reaktor pirolisis yang terbuat dari stainless steel berbentuk silinder dengan volume 0,96 m3 dengan temperatur akhir 450oC. hasil penelitian menunjukkan bahwa semakin banyak penambahan Plastik berlapisan aluminium foil maka semakin cepat naiknya temperatur mencapai titik optimum yang ditetapkan (450oC. Sedangkan senyawa kimia yang dihasilkan pada pirolisis yang mengandung PS sebagian besar berupa senyawa aromatic, sedangkan pada pirolisis AL sebagian besar berupa senyawa olefin.

  1. Opinion about difficulties of RA reactor operation under conditions of high activity of the heavy water system - Annex 2; Prilog 2 - Misljenje o teskocama eksploatacije reaktora RA u uslovima visoke aktivnosti teskovodnog sistema

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    It was concluded that reactor the reactor operation is very dangerous for the reactor installation as well as safety of the staff under conditions of heavy water increased activity. Two fundamental arguments in favour of this conclusion are: insufficient possibility of reactor components inspection during maintenance and operation in the future period; difficulties in prevention of accidents that could occur is equally dangerous for the reactor facility and the environment. Cleaning and decontamination of the complete heavy water system is needed before the reactor operation starts in order to avoid possible failures or accidental events. [Serbo-Croat] Zakljuceno je da je eksploatacija reakora u uslovima postojece aktivnosti teskovodnog sistema veoma opasna po sam reaktor i po personal. Dva osnovna razloga u prilog ovog zakljucka su: nedovoljna mogucnost kontrole ispravnosti svih elemenata reaktora u toku remonta i ekspolatacije u predstojecem periodu; teskoca borbe sa udesima, koji bi se eventualno dogodili podjednako je opasna po instalaciju i okolinu. Pre no sto se nastavi sa daljom ekspolatacijom reaktora potrebno ciscenje, dekontaminacija sistema teske vode kako bi se izbegla moguca ostecenja ili akcidentalne situacije.

  2. The physics design of EBR-II; Physique du reacteur EBR-II; Fizicheskij raschet ehksperimental'nogo reaktora - razmnozhitelya EVR-II; Aspectos fisicos del reactor EBR-II

    Energy Technology Data Exchange (ETDEWEB)

    Loewenstein, W. B. [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    hexagonal del reactor los datos obtenidos en geometrias simples idealizadas, analiticas o experimentales. Se compara el rendimiento nuclear, incluso el de reproduccion, del reactor real con el del modelo teorico y se describen las variaciones a largo plazo de la reactividad y de la generacion de energia en la envoltura fertil, refiriendolas a los ciclos propuestos para el combustible y la envoltura fertil. La memoria formula consideraciones sobre la seguridad estudiando en particular la introduccion de indices de reactividad normales y anormales y la consecuencia de supuestos efectos de reactividad, que se basan en el comportamiento fisico de la aleacion combustible y de la estructura del reactor, asi como en la extrapolacion al sistema del EBR-II de los experimentos realizados con el conjunto TREAT. Por ultimo, examina el problema de la fusion del cuerpo del reactor EBR-II. (author) [Russian] Vychisleniya statisticheskogo, dinamicheskogo i dlitel'nogo rezhima reaktivnosti ehksperimental'nog o reaktora-razmnozhitel ya EBR-II dayutsya sovmestno s rezul'tatami i analizom ehksperimentov na reaktore EBR-II po dostizheniyu kritichnosti v sukhom sostoyanii i otdel'nykh ehksperimentov na reaktore ZPR-III. Osoboe vnimanie udelyaetsya problemam fizicheskogo rascheta reaktora, kotorye voznikayut posle opredeleniya printsipial'noj skhemy i do sooruzheniya ili vvoda reaktora v ehkspluatatsiyu. Opisyvayutsya analiz bezopasnosti reaktora i soobrazheniya po otsenke opasnostej, a takzheikh vliyanie na raschet reaktora. V doklade opisyvaetsya sposob ispol'zovaniya modeli EBR-II na osnovanii poluchennykh na reaktore ZPR-III dannykh, a takzhe dannykh sukhoj kritichnosti reaktora EBR-II. EHti ehksperimenty, ikh analiz i teoreticheskie vykladki yavlyayutsya osnovoj dlya opredeleniya fizicheskogo povedeniya reaktornoj sistemy. Bolee podrobno issleduyutsya ogranicheniya, prisushchie primeneniyu ehksperimental'nykh dannykh k rassmatrivaemo j ehnergeticheskoj reaktornoj sisteme. Syuda otnosyatsya

  3. An advanced three-dimensional simulation system for safety analysis of gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lapins, Janis

    2016-07-01

    simulated simultaneously. TORT-TD and ATTICA{sup 3D} exchange data (power distributions or fuel and moderator temperature distribution, possibly hydrogen distribution) by means of a common interface that interpolates values that are exchanged on mutual computational grids by volumetric averaging. As verification for the proper operation of the interface, the steady state of the transient PBMR-400 benchmark was used. After obtaining a coupled steady state, the transient exercises are performed to test the proper working of the interface in time dependent cases. Here, the cold helium ingress, the total control rod withdrawal case and the total control rod ejection case were simulated and compared to results of other partakers of the benchmark. Also, the coupled system was validated for a full power temperature distribution experiment in the Chinese experimental reactor HTR-10 where good agreement could be reached with the measurements. The coupled HTR simulation system TORT-TD/ATTICA{sup 3D} was then applied for single control rod ejection cases for both the PBMR-400 and the HTR-PM. These cases require a 180 model of the reactor. As preparatory works, the control rod cross sections were adjusted to yield the same reactivity increase as the grey curtain model for the PBMR and with MCNP5 for the HTR-PM. Since there are strong shielding effects by neighbouring rods, the power increase was moderate due to strong Doppler and moderator feedbacks. For the HTR-PM, coupled calculations for water ingress cases are simulated. This also tested the whole computational sequence, i.e. steam transport into the core by ATTICA{sup 3D}, then transfer of hydrogen densities (from hydrogen or from steam) to TORT-TD via the interface, interpolation of the macroscopic cross sections which changes the power density, and the feedback to ATTICA{sup 3D}. Additionally, an anticipated transient without scram is simulated where shutdown of the reactor is achieved by the temperature feedback effects. For

  4. Prismatic Core Coupled Transient Benchmark

    International Nuclear Information System (INIS)

    Ortensi, J.; Pope, M.A.; Strydom, G.; Sen, R.S.; DeHart, M.D.; Gougar, H.D.; Ellis, C.; Baxter, A.; Seker, V.; Downar, T.J.; Vierow, K.; Ivanov, K.

    2011-01-01

    The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The benchmark-working group is currently seeking OECD/NEA sponsorship. This benchmark is being pursued and is heavily based on the success of the PBMR-400 exercise.

  5. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  6. Cross-section parameterization of the pebble bed modular reactor using the dimension-wise expansion model

    International Nuclear Information System (INIS)

    Zivanovic, Rastko; Bokov, Pavel M.

    2010-01-01

    This paper discusses the use of the dimension-wise expansion model for cross-section parameterization. The components of the model were approximated with tensor products of orthogonal polynomials. As we demonstrate, the model for a specific cross-section can be built in a systematic way directly from data without any a priori knowledge of its structure. The methodology is able to construct a finite basis of orthogonal polynomials that is required to approximate a cross-section with pre-specified accuracy. The methodology includes a global sensitivity analysis that indicates irrelevant state parameters which can be excluded from the model without compromising the accuracy of the approximation and without repetition of the fitting process. To fit the dimension-wise expansion model, Randomised Quasi-Monte-Carlo Integration and Sparse Grid Integration methods were used. To test the parameterization methods with different integrations embedded we have used the OECD PBMR 400 MW benchmark problem. It has been shown in this paper that the Sparse Grid Integration achieves pre-specified accuracy with a significantly (up to 1-2 orders of magnitude) smaller number of samples compared to Randomised Quasi-Monte-Carlo Integration.

  7. Evaluation of the Control Rod Super Alloy Material of HTR-PM

    International Nuclear Information System (INIS)

    Li Pengjun; Yan He; Diao Xingzhong

    2014-01-01

    The control rod drive mechanism (CRDM) system is served as the first reactivity control and shutdown system for the high temperature reactor pebble-bed module (HTR-PM) in Shandong, China. And the control rod, which is pulled up and down by a chain sprocket mechanism of CRDM to realize reactivity control, compensation and shutdown, has to be durable under temperature as high as 550℃ for a long time. Thus the material persistent strength under high temperature is quite important for the reliability of the CRDM. In this paper, a review on material selection of control rod of high temperature gas cooled reactors, including AVR and THTR-300 in Germany, HTTR in Japan, PBMR in South Africa and Dragon in Britain, was summarized. The major parameters of two kinds of high temperature alloy, incoloy 800H and alloy 625, were compared and discussed. According to the ASME NH volume, a design criterion for the control rod was established and applied in the analysis of the chain by using finite element method. The numerical simulations showed that the chain made of alloy 625 could meet the condition and work for a long time under high temperature. (author)

  8. An endothermic chemical process facility coupled to a high temperature reactor. Part II: Transient simulation of accident scenarios within the chemical plant

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Revankar, Shripad T.

    2012-01-01

    Highlights: ► Seven quantitative transient case studies were analyzed in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Positive power excursion in the nuclear reactor were found for helium-inlet overcoolings. ► In all cases studied the maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Transient study of the operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven quantitative transient case studies are analyzed. The case studies consist of: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. Various parametric studies based on the magnitude of the events were also performed. The only chemical plant initiated events that caused a positive power excursion in the nuclear reactor were helium-inlet overcoolings due to process holding tank failures or reaction chamber ruptures. Even for a severe sustained overcooling, the calculated maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. The qualitative basis for the case studies and the analysis models are summarized in part I of this paper.

  9. RA reactor operation in 1991, Part 1; Deo 1 - Pogon i odrzavanje nuklearnog reaktora RA u 1991. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Cupac, S; Sulem, B; Zivotic, Z; Majstorovic, D; Sanovic, V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1991, three major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. Renewal of the reactor instrumentation was started but but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA for political reasons. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Training of the existing personnel was done regularly, but lack of financial support prevented employment of new workers that would be needed for operation in shifts and regular maintenance. [Serbo-Croat] U proteklom periodu reaktor RA nije bio u pogonu zato sto je 30. jula Republicki komitet za zdravlje i socijalnu politiku republike Srbije, zabranio njegov rad zbog toga sto reaktor ne poseduje sistem za udesno hladjenje i ne poseduje odgovarajuce filtere u sistemu specijalne ventilacije. Kako bi se ubuduce mogao obezbediti

  10. RB research nuclear reactor, Annual report for 1981; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Sotic, O; Pesic, M; Vranic, S; Zivkovic, B; Bogdanovic, M; Petronijevic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1981-07-01

    The annual report for 1981 includes the following: utilization of the RB reactor; accident and incidents analysis; description of the reactor equipment status; dosimetry and radiation protection; RB reactor staff; financial data. Seven Annexes to this report are concerned with: maintenance of the reactor components and equipment, including nuclear fuel, heavy water, reactor vessel, heavy water coolant circuit, experimental platforms, absorption rods; maintenance of the electric power supply system, neutron source equipment, crane; control and maintenance of ventilation and heating systems, gas and comprised gas systems, fire protection system; plan for renewal of the reactor components; contents of the RB reactor safety report; reactor staff; review of measured radiation doses; experimental methods; training of the staff; and financial report.

  11. RB research nuclear reactor, Annual report for 1982; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1982-12-15

    This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 7 Annexes as follows: Maintenance of reactor equipment in 1982; contents of the RB reactor safety report; review of radiation doses in the reactor building and exposure doses for the reactor staff; utilization of the RB reactor in 1982; and financial data.

  12. Neutronic modeling of pebble bed reactors in APOLLO2

    International Nuclear Information System (INIS)

    Grimod, M.

    2010-01-01

    In this thesis we develop a new iterative homogenization technique for pebble bed reactors, based on a 'macro-stochastic' transport approximation in the collision probability method. A model has been developed to deal with the stochastic distribution of pebbles with different burnup in the core, considering spectral differences in homogenization and depletion calculations. This is generally not done in the codes presently used for pebble bed analyses, where a pebble with average isotopic composition is considered to perform the cell calculation. Also an iterative core calculation scheme has been set up, where the low-order RZ S N full-core calculation computes the entering currents in the spectrum zones subdividing the core. These currents, together with the core k eff , are then used as surface source in the fine-group heterogeneous calculation of the multi-pebble geometries. The developed method has been verified using reference Monte Carlo simulations of a simplified PBMR- 400 model. The pebbles in this model are individually positioned and have different randomly assigned burnup values. The APOLLO2 developed method matches the reference core k eff within ± 100 pcm, with relative differences on the production shape factors within ± 4%, and maximum discrepancy of 3% at the hotspot. Moreover, the first criticality experiment of the HTR-10 reactor was used to perform a first validation of the developed model. The computed critical number of pebbles to be loaded in the core is very close to the experimental value of 16890, only 77 pebbles less. A method to calculate the equilibrium reactor state was also developed and applied to analyze the simplified PBMR-400 model loaded with different fuel types (UO 2 , Pu, Pu + MA). The potential of the APOLLO2 method to compute different fluxes for the different pebble types of a multi-pebble geometry was used to evaluate the bias committed by the average composition pebble approximation. Thanks to a 'compensation of error

  13. Current status and results of the PBMR -Pebble Box- benchmark within the framework of the IAEA CRP5 - 341

    International Nuclear Information System (INIS)

    Reitsma, F.; Tyobeka, B.

    2010-01-01

    The verification and validation of computer codes used in the analysis of high temperature gas cooled pebble bed reactor systems has not been an easy goal to achieve. A limited amount of tests and operating reactor measurements are available. Code-to-code comparisons for realistic pebble bed reactor designs often exhibit differences that are difficult to explain and are often blamed on the complexity of the core models or the variety of analysis methods and cross section data sets employed. For this reason, within the framework of the IAEA CRP5, the 'Pebble Box' benchmark was formulated as a simple way to compare various treatments of neutronics phenomena. The problem is comprised of six test cases which were designed to investigate the treatments and effects of leakage and heterogeneity. This paper presents the preliminary results of the benchmark exercise as received during the CRP and suggests possible future steps towards the resolution of discrepancies between the results. Although few participants took part in the benchmarking exercise, the results presented here show that there is still a need for further evaluation and in-depth understanding in order to build the confidence that all the different methods, codes and cross-section data sets have the capability to handle the various neutronics effects for such systems. (authors)

  14. Pengaruh Pretreatment Secara Alkalisasi-Resistive Heating terhadap Kandungan Lignoselulosa Jerami Padi

    Directory of Open Access Journals (Sweden)

    Dewi Maya Maharani

    2017-09-01

    Full Text Available Cellulose is a potential biomass that is used for bioethanol production and commonly present in agricultural residues like rice straw. Cellulose is an important material to produce glucose and bioethanol, but it is covered by lignin and hemicellulose bonds to form a lignocellulose.  Bioethanol production using basic material containing cellulose requires special attention in the process of pretreatment for lignin degradation process and increase the accessible surface and decrystallize cellulose. The aim of this research was to apply alkalization and resistive heating combine method for rice straw pretreatment process before further being converted into bioethanol and to determine the effects of heating temperature and NaOH concentration on the content of  lignin, cellulose, and hemicellulose. The reactor had been designed for resistive heating process. Rice straw that was resized into 100 mesh has dissolved with 0.03 M, 0.05 M, and 0.07 M NaOH and heated with resistive heating temperature of 75 oC, 85 oC, and 99 oC. Cellulose is a raw material that will be further converted into glucose. So that, the selected optimum conditions of this study were  pretreatment with the highest increase of cellulose content level until 8.88% and resulted decreasing levels of lignin (1.39% and hemicellulose (4.33% by temperature  75 oC and 0.07 M NaOH concentration. Resistive heating that combine with alkalization can be used for rice straw pretreatment process that reduce lignin and hemicellulose content as well as increasing cellulose content.   ABSTRAK Selulosa merupakan biomassa yang potensial digunakan untuk produksi bioetanol dan banyak ditemukan di residu pertanian seperti jerami padi. Selulosa merupakan material penting yang dapat dikonversi menjadi glukosa kemudian dikonversi menjadi bioetanol, namun selulosa pada alam dilapisi oleh ikatan lignin dan hemiselulosa menjadi lignoselulosa. Pembuatan bioetanol berbasis selulosa membutuhkan proses

  15. The effective lifetime and temperature coefficient in a coupled fast-thermal reactor; Temps de vie effectif et coefficient de temperature dans un reacteur a couplage neutrons rapides-neutrons thermiques; Ehffektivnyj srok zhizni i temperaturnyj koehffitsient nejtronov v dvoyakom reaktore na bystrykh i teplovykh nejtronakh; Vida efectiva y coeficiente de temperatura en un reactor con acoplamiento rapido-termico

    Energy Technology Data Exchange (ETDEWEB)

    Haefele, W. [Kernforschungszentrum, Karlsruhe (Germany)

    1962-03-15

    The theory of coupled systems was extensively developed by Avery and co-workers at the Argonne National Laboratory. One of the main points of interest in a coupled system is the larger effective lifetime of neutrons. The effect of the thermal component acts as a sort of neutron-delayer. As in the theory of delayed neutrons the delaying effect disappears if the reactivity worth is high enough to make the fast component critical by itself. In the study a coupled reactor is considered where the fast component suffers a sudden reactivity step {alpha}{sub 0}. Because of the increasing power-level the temperature rises and two temperature coefficients start to work: the temperature coefficient of the fast component and the temperature coefficient of the thermal component. The problem is considered with one group of delayed neutrons (in the ordinary meaning). A formalism is given to express the effective lifetime and temperature coefficient during the different stages of the excursion. Excursions for different {alpha}{sub 0} are given so that the limit of fast-reactor kinetics is reached. (author) [French] La theorie des systemes a couplage a ete mise au point par Avery et ses collaborateurs au Laboratoire national d'Argonne. L'une des caracteristique les plus interessantes d'un systeme a couplage est que le temps de vie effectif des neutrons est plus long. L'effet de la partie thermique contribue en quelque sorte a retarder les neutrons. Comme dans la theorie des neutrons retardes, l'effet de retardement disparait lorsque la reactivite a une valeur suffisamment elevee pour rendre la partie rapide critique par elle-meme. L'auteur du memoire considere un reacteur a couplage dont la partie rapide subit un saut instantane de reactivite, {alpha}{sub 0}. La temperature s'eleve a cause de l'augmentation de puissance et deux coefficients de temperature commencent a s'appliquer: le coefficient de temperature de la partie rapide et le coefficient de temperature de la partie

  16. Research reactor RB, technical characteristics and experimental possibilities; Zbornik radova, Konferencija o koriscenju nuklearnih reaktora u Jugoslaviji

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Vranic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1978-05-15

    Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) Istrazivacki nuklearni reaktor RB u Laboratoriji za nuklearnu energetiku i tehnicku fiziku Instituta za nuklearne nauke 'Boris Kidric' u Vinci je prvi reaktorski sistem izgradjen u Jugooslaviji 1958. godine. U ovom radu opisane su osnovne tehnicke karakteristike tog reaktora, kao i mogucnosti za izvodjenje eksperimenata koje on pruza korisnicima. Njegova relativno jednostavna konstrukcija i fleksibilnost omogucavaju da se na njemu izvrse direktna merenja niza fizickih parametara, a s druge strane odsustvo bioloskog zastitnog omotaca cini ga veoma pogodnim za razna bioloska i druga ozracivanja, a takodje i dozimetrijska merenja gde se zahteva snazan izvor neutrona. (author)

  17. PEMBUATAN KALSIUM KARBONAT DARI BITTERN DAN GAS KARBON DIOKSIDA SECARA KONTINYU

    Directory of Open Access Journals (Sweden)

    Soemargono Soemargono

    2012-01-01

    Full Text Available Kalsium karbonat  yang  digunakan  dalam  industri- industri cat, karet, dan  kertas  harus  mempunyai  mutu yang  tinggi, terutama  kemurnian  dan kehalusannya.Untuk itu, Indonesia masih  mendatangkan  kalsium  karbonat murni dari luar negeri dalam jumlah yang cukup besar. Bittern merupakan bahan buangan industri garam yang disebut juga air tua, mengandung senyawa kalsium. Karbon dioksida biasanya berasal dari hasil pembakaran yang masuk ke udara. Kandungannya di udara kecil, tetapi berpotensi sebagai pencemar. Dengan mereaksikan kalsium yang terkandung dalam bittern dengan gas CO2 akan terbentuk CaCO3 dalam suasana basa. Pembentukan kalsium karbonat dilakukan dengan proses kontinyu dalam reaktor kolom bersekat miring. Dari hasil penelitian diperoleh bahwa pengendapan magnesium dengan larutan ammonia menyebabkan kandungan kalsium ikut terdegradasi. Hasil terbaik yang diperoleh dicapai pada kondisi pH awal, kecepatan alir gas CO2, kecepatan alir cairan, dan suhu masing-masing pada 8,7; 2265 mL/menit; 10 mL/menit; dan 303 K, dengan konversi sebesar 38,40%. Produk berupa CaCO3, yang diperoleh mempunyai kemurnian sebesar 21,34%.

  18. CRITICALITY ANALYSIS OF URANIUM STORAGE FACILITY WITH FORMATION RACKS

    Directory of Open Access Journals (Sweden)

    Sri Kuntjoro

    2017-03-01

    ANALISIS KRITIKALITAS DI FASILITAS PENYIMPANAN BAHAN URANIUM DENGAN FORMASI PENGATURAN RAK. Bahan uranium dibutuhkan untuk produksi bahan bakar reaktor penelitian dan radioisotop. Bahan uranium sebelum digunakan terlebih dahulu disimpan pada fasilitas penyimpanan. Salah satu prasyarat fasilitas penyimpanan bahan uranium adalah fasilitas tersebut harus dalam kondisi sub-kritis. Bila kondisi kritis terjadi mengakibatkan proses fissi pada bahan uranium tidak terkendali, sehingga akan menimbulkan suhu yang sangat tinggi. Tujuan dari penelitian ini adalah untuk menganalisa kondisi kritikalitas dari fasilitas penyimpanan bahan uranium yang berada di PT. INUKI (Persero untuk menjamin fasilitas tersebut dalam kondisi sub-kritis. Analisis kritikalitas dilakukan menggunakan program MCNP-5 untuk mengetahui tingkat kritikalitas dari tiga fasilitas penyimpanan bahan uranium untuk kondisi awal dan kondisi setelah ditambahkan rak penyimpanan. Untuk fasilitas penyimpanan 1 dan 2 dibuat tiga skenario pengaturan container pada rak penyimpanan, sedangkan pada fasilitas penyimpanan 3 dilakukan 1 skenario.  Hasil ini menunjukkan seluruh fasilitas penyimpanan pada kondisi awal dan setelah ditambah rak penyimpanan dalam kondisi sub-kritis (k-eff<1. Hasil tersebut selanjutnya dipergunakan sebagai dasar untuk menyusun manejemen pengelolaan bahan uranium. Selain itu juga digunakan sebagai dasar untuk pembuatan ijin dari badan pengawas (BAPETEN. Kata Kunci : kritikalitas, fasilitas penyimpanan berbahan uranium,  k-eff

  19. Methods for measuring nuclear properties of materials, Safety coefficient method and measurement of effective absorption coefficient of graphite by safety coefficient method; Razvijanje metoda merenja nuklearnih karakteristika materijala, Razrada metode koeficijenta opasnosti i merenje efektivnog apsorpcionog preseka grafita metodom koeficijenta opasnosti

    Energy Technology Data Exchange (ETDEWEB)

    Maglic, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1962-11-15

    Reactivity of a reactor depends on production, absorption and leaking of neutrons. Change of absorption causes reactivity change, and this fact is used for determining the neutron absorption cross section for the sample inserted in the reactor core. Method for determining the absorption cross section based on reactivity change is called method of safety coefficient. Measurements of neutron absorption cross section for graphite was done in the RA reactor vertical experimental channel VK-5. taking into account the results obtained for five types of graphite this method is considered to be reliable for use. Comparison of nuclear properties of different types of graphite was done as well. Reaktivnost reaktora zavisi od proizvodnje neutrona, apsorpcije i isticanja neutrona. Promena apsorpcije izaziva promene reaktivnosti reaktora pa se ova osobina koristi za odedjivanje neutronskog apsorpcionog preseka uzorka koji se unosi u reaktor. Metoda merenja apsorpcionog preseka na bazi promene reaktivnosti nazvana je metodom koeficijenta opasnosti. Merenje apsorpcionog preseka grafita uradjeno je na reaktoru RA u vertikalnom eksperimentalnom kanalu VK-5. S obzirom na rezultate koji su dobijeni za pet vrsta grafita moze se smatrati da je opravdano koriscenje ove metode. Izvrseno je i poredjenje nuklearnih osobina pomenutih tipova grafita.

  20. Impact Of Secondary-Primary Pumps Operating Sequence On The Electrical Power Supply System

    International Nuclear Information System (INIS)

    Suwoto; Rusdiyanto; Kiswanto

    2001-01-01

    The operating procedure of the reactor cooling system has decided that the primary cooling pump should be operated before secondary cooling pump as known primary-secondary pumps operating sequence. This decision is based on consideration that starting current of the primary pump is higher than secondary pump. Therefore, the primary-secondary pumps operating sequence can avoid the power supply system failure. However, this operating procedure has to take a consequence that in case of primary pump failure, the shutdown time period of the reaktor to be longer caused to re operate the primary pump has required that the running secondary pump should be shutted off. To solve this problem, an impact analysis of the secondary-primary pumps operating sequence on the electric power supply system was carried out to identify the revision possibility of the cooling pump operating procedure. The analysis by discussion of the measuring results of the secondary and primary pump starting current related to another electrical loads has been measured. From discussion it can be concluded that secondary-primary pumps operating sequence has no impact to failure in electric power supply system

  1. Pengaruh penambahan Ca(OH2 pada Proses Pirolisis terhadap Hasil Gasifikasi Batubara Bituminus dengan medium Gas CO2

    Directory of Open Access Journals (Sweden)

    Saripah Sobah

    2014-06-01

    Full Text Available Pemanfaatan batubara melalui proses gasifikasi perlu dikembangkan lebih lanjut karena proses ini dapat dijadikan alternatif untuk menggantikan peranan  gas alam sebagai sumber gas sintesis. Di samping itu, proses ini dapat mengurangi pencemaran lingkungan karena teknologi gasifikasi merupakan teknologi yang bersih dan dapat mengurangi jumlah gas CO2 yang dibuang ke lingkungan. Penelitian ini bertujuan untuk mengetahui pengaruh penambahan Ca(OH2 pada proses pirolisis terhadap hasil gasifikasi arang batubara bituminus dengan medium gas CO2. Reaksi karbon dari arang batubara dengan gas CO2 pada proses gasifikasi merupakan reaksi endotermis dan berlangsung sangat lambat pada suhu di bawah 1000oC sehingga digunakan Ca(OH2 sebagai katalisator. Proses gasifikasi batubara dijalankan dalam reaktor fixed bed. Hasil penelitian menunjukkan bahwa gasifikasi arang batubara dengan penambahan Ca(OH2 pada proses pirolisis memberikan pengaruh terhadap komposisi gas hasil yaitu berkurangnya kadar gas CO2 dan menyebabkan berkurangnya kadar belerang pada arang hasil pirolisis dan gasifikasi. Proses ini juga dapat mengurangi kadar gas CO2 sebesar 63,17% dan untuk  gasifikasi tanpa Ca(OH2 , CO2 dapat dikurangi kadarnya sampai 35,2%.

  2. Conceptual Design of On-line Based Licensing Review and Assesment System of Nuclear Installations and Nuclear Materials ('PRIBEN')

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2008-01-01

    At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. There are only three research reactors, one nuclear fuel fabrication plant for research reactors, and one experimental fuel fabrication plant for nuclear power, one isotope production facility and some other research facilities. All the facility is under Nuclear Energy Regulatory Agency (BAPETEN) controlling through regulation, licensing and inspection. The organizations operation submits licensing application to BAPETEN before utilizing the facility. According to the regulation before BAPETEN give license they perform review and assessment for the utility application. Based on the review and assessment result, BAPETEN may stipulate, reject, delay or terminate the license. In anticipation of expansion of the nuclear program in Indonesia, BAPETEN should have an integrated and updated system for review and asses the licensing application. For this reason, an expert system for the review and asses the licensing application, so-called PRIBEN (Perizinan Reaktor, Instalasi dan Bahan Nuklir/Licensing of Reactor, Nuclear Installations and Nuclear Materials), is developed which runs on the online-based reality environment

  3. An HTR cogeneration system for industrial applications

    International Nuclear Information System (INIS)

    Haverkate, B.R.W.; Heek, A.I. van; Kikstra, J.F.

    2001-01-01

    Because of its favourable characteristics of safety and simplicity the high-temperature reactor (HTR) could become a competitive heat source for a cogeneration unit. The Netherlands is a world leading country in the field of cogeneration. As nuclear energy remains an option for the medium and long term in this country, systems for nuclear cogeneration should be explored and developed. Hence, ECN Nuclear Research is developing a conceptual design of an HTR for Combined generation of Heat and Power (CHP) for the industry in and outside the Netherlands. The design of this small CHP-unit for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. The concept that was subject of this study, INCOGEN, used a 40 MW thermal pebble bed HTR and produced a maximum amount of electricity plus low temperature heat. The system has been improved to produce industrial quality heat, and has been renamed ACACIA. The output of this installation is 14 MW electricity and 17 tonnes of steam per hour, with a pressure of 10 bar and a temperature of 220 deg. C. The economic characteristics of this installation turned out to be much more favourable using modern data. The research work for this installation is embedded in a programme that has links to the major HTR projects in the world. Accordingly ECN participates in several IAEA Co-ordinated Research Programmes (CRPs). Besides this, ECN is involved in the South African PBMR-project. Finally, ECN participates in the European Concerted Action on Innovative HTR. (author)

  4. Strategy of VHTR Realization

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    High temperature gas cooled reactor has been developed since 1956. Fundamental idea of a gas cooled reactor is to achieve high temperature which is suitable for high efficiency application such as electricity generation. The core is composed of ceramics, graphite blocks which are mechanical stable up to very high temperature. Fuel is ceramics, TRISO ( tri-isotropic coated micro particle) whose dense coating layers work as small radioactivity containment. Coolant is inert gas, helium, which is stable chemically, neutronically, and thermal hydraulically. Several test reactors such as DRE, PB-1, FSV, AVR, THTR, HTTR, HTR-10 were built and demonstrated their safety. Large GA-HTR, RSA-PBMR projects are canceled and US-NGNP project is idling. Only Chinese HTR-PM demonstrator is under construction. HTGR has long history of development. For realization and market penetration, VHTR community should look at niche market such as carbon free energy supply to industry complex, electric power for small grid, carbon free hydrogen production, power source for space colony. Technology Readiness Level should be advanced to get proper investment from industry. For this, cooperation between international R and D institutions is required. Clearly divided role between universities, research institutions, and industries will reduce complication and shorten VHTR realization day.

  5. Development of the South African Network for Nuclear Education, Science and Technology

    International Nuclear Information System (INIS)

    Cilliers, A.

    2016-01-01

    Full text: South Africa has long been regarded as an active country in the nuclear industry with two operating power reactors and a research reactor. In recent years’ research and development projects, such as the Pebble Bed Modular Reactor, has established additional expertise in the country situated at various institutions. After the PBMR project was stopped, the expertise became fragmented throughout the country and some experts even left the country. A number of training and research facilities have also completed their research cycle and are in the process of being decommissioned. With the renewed interest in nuclear technology and the states position to complete the procurement of 9600 MW of nuclear power before the end of the year, nuclear knowledge gap has been identified and the need to capture all nuclear education and research in an educational network as well as to establish new nuclear training and research facilities such as small training reactors and research laboratories to support the national new build programme. This expertise and research facilities were combined into SAN-NEST (South African Network for Nuclear Education, Science and Technology) for South Africa and the African continent, with links to AFRA-NEST. The paper reports on the successes and challenges of the establishment and operation of SAN-NEST. (author

  6. Strategy of VHTR Realization

    International Nuclear Information System (INIS)

    Chang, Jonghwa

    2015-01-01

    High temperature gas cooled reactor has been developed since 1956. Fundamental idea of a gas cooled reactor is to achieve high temperature which is suitable for high efficiency application such as electricity generation. The core is composed of ceramics, graphite blocks which are mechanical stable up to very high temperature. Fuel is ceramics, TRISO ( tri-isotropic coated micro particle) whose dense coating layers work as small radioactivity containment. Coolant is inert gas, helium, which is stable chemically, neutronically, and thermal hydraulically. Several test reactors such as DRE, PB-1, FSV, AVR, THTR, HTTR, HTR-10 were built and demonstrated their safety. Large GA-HTR, RSA-PBMR projects are canceled and US-NGNP project is idling. Only Chinese HTR-PM demonstrator is under construction. HTGR has long history of development. For realization and market penetration, VHTR community should look at niche market such as carbon free energy supply to industry complex, electric power for small grid, carbon free hydrogen production, power source for space colony. Technology Readiness Level should be advanced to get proper investment from industry. For this, cooperation between international R and D institutions is required. Clearly divided role between universities, research institutions, and industries will reduce complication and shorten VHTR realization day

  7. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  8. Introducing nuclear power into currently non-nuclear states

    International Nuclear Information System (INIS)

    Claassen, Gert

    2007-01-01

    As the nuclear renaissance gains momentum, many countries that currently have no nuclear power plants will begin to consider introducing them. It is anticipated that smaller reactors such as the Pebble Bed Modulator Reactor (PBMR) will not only be sold to current nuclear states to also to states where there is currently no nuclear experience. A range of issues would have to be considered for nuclear plants to be solid to non-nuclear states, such as the appropriate regulatory environment, standardization and codes, non-proliferation, security of supply, obtaining experienced merchant operators, appropriate financial structures and education and training. The paper considers nine major issues that need to be addressed by governments and vendors alike. International cooperation by organisations such as the IAEA, financial institutions and international suppliers will be required to ensure that developing countries as well as developed ones share the benefits of the nuclear renaissance. The opportunities that the nuclear industry affords to develop local skills, create job opportunities and to develop local manufacturing industries are among the important reasons that the South African Government has decided to support and fund the development of the Pebble Bed Modular Reactor project. These considerations are included in the paper. (author)

  9. Compact heat exchanger technologies for the HTRs recuperator application

    International Nuclear Information System (INIS)

    Thonon, B.; Breuil, E.

    2001-01-01

    Modern HTR nuclear power plants which are now under development (projects GT-MHR, PBMR) are based on the direct cycle concept. This concept leads to a more important efficiency compared to the steam cycle but requires the use of high performance components such as an helium/helium heat exchanger called recuperator to guarantee the cycle efficiency. Using this concept, a net plant efficiency of around 50% can be achieved in the case of an electricity generating plant. As geometric constraints are particularly important for such a gas reactor to limit the size of the primary vessels, compact heat exchangers operating at high pressure and high temperature are attractive potential solutions for the recuperator application. In this frame, Framatome and CEA have reviewed the various technologies of compact heat exchangers used in industry. The first part of the paper will give a short description of the heat exchangers technologies and their ranges of application. In a second part, a selection of potential compact heat exchangers technologies are proposed for the recuperator application. This selection will be based upon their capabilities to cope with the operating conditions parameters (pressure, temperature, flow rate) and with other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. (author)

  10. Korea-South Africa nuclear energy focal point

    International Nuclear Information System (INIS)

    Kim, Y. H.; Lee, K. H.; Park, M. K.; Chae, J. S.; Cheon, K. J.

    2007-09-01

    The major objective of the project is to hold Korea-S. Africa joint nuclear or workshop and establish a network of nuclear experts between the two countries. In addition, identification of the collaboration items is another objective of this project. the outcome of this program would be a cornerstone for the nuclear collaboration between Korea and S. Africa. African countries is usually very abundant in the natural resources and developed countries are trying to improve relationships with African countries to secure the natural resources, S. Africa is also very rich with some resources including uranium. Therefore, Korea depending on foreign resources needs to improve the relationship with S. Africa form the viewpoint of securing natural resources. S. Africa has two PWR-type nuclear power plants and is planning to deploy more PWRs to increase the nuclear capacity. In addition, S. Africa is a leading country in developing the PBMR, which is a type of Generation-IV VHTR (very high temperature reactor). In Korea, the VHTR technologies are also under development for the nuclear hydrogen. Although Korea has its own PWR brand, it is not well known to S. Africa. The focal point program would lay a cornerstone for mutually beneficial nuclear collaborations between two countries

  11. Passive device for emergency core cooling of pressurized water reactors. Pasivno ustrojstvo za bezopasnost na vodo-voden atomen reaktor

    Energy Technology Data Exchange (ETDEWEB)

    Sikora, D

    1984-02-28

    The device proposed ensures additional margin of reactor subcriticality in case of post-accident emergency core cooling (ECC), using concentrated solution of chemical absorber and hot water from the secondary circuit. It consists of: a) a differential cylinder with a differential piston in it, with a lid and a seal, connected to a pipeline for secondary coolant; b) a pipeline for the secondary coolant; c) a volume between the lid and the piston for the secondary coolant from the steam generator; d) a discharge pipeline with a check valve of seal type connecting the inner volume of the differential cylinder to the discharge line; and e) a pipeline from the high-pressure volume of the differential cylinder filled with concentrated chemical absorber solution, to one of the main circulation loops. The device permits ECC innovation of the operating non-standard nuclear power plants with PWR type reactors.

  12. Possibilities of using metal uranium fuel in heavy water reactors; Mogucnosti upotrebe metalnog urana kao goriva za teskovodne reaktore

    Energy Technology Data Exchange (ETDEWEB)

    Djuric, B; Mihajlovic, A; Drobnjak, Dj [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions. Postoje ozbiljni ekonomski razlozi za primenu metalnog urana u teskovodnim reaktorima, pre svega zbog njegove velike gustine, odnosno visokog konverzionog faktora, i zbog niskih troskova proizvodnje gorivnih elemenata. Glavne prepreke su bubrenje pri velikim stepenima sagorevanja i opasnost od korozije. Postoje veliki izgledi da se primenom odredjenih projektnih koncepcija i upotrebom legiranjem poboljsanog urana postigne zadovoljavajuca stabilnost metalnog urana u uslovima rada reaktora (author)

  13. Report on the interpretation of critical experiments in the Siemens-Argonaut-Reactor Graz to study water ingress into spherical elements. Ergebnisbericht zur Auslegung kritischer Experimente am Siemens-Argonaut-Reaktor Graz zum Studium des Wassereinbruches im Kugelhaufen

    Energy Technology Data Exchange (ETDEWEB)

    Schuerrer, F [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik; Neef, R D [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

    1979-04-15

    The experiments described are of interest in the study of water contamination in HTR fuel elements. The Siemens Argonaut Reactor (SAR) has been considered as a research tool for a simulation experiment. Following a brief description of the SAR, planned programs are discussed in 'dry' and 'wet' cores. Detector foil types and locations are noted. A theoretical model is developed and nuclide concentrations estimated in the various spectral zones. Reactivity calculations have been made and are summarised for various H{sub 2}O percentage concentrations. The discussion is supported by simplified core layout diagrams and graphs of core flux distributions. Neutron diffusion and spectra calculations are referenced to computer programs used by KFA-Juelich, published elsewhere, and include GAM, THERMOS, MUPO and EXTERMINATOR-2. (G.C.)

  14. MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR

    Directory of Open Access Journals (Sweden)

    Suwardi Suwardi

    2015-03-01

    Full Text Available MODEL SIMULATION OF GEOMETRY AND STRESS-STRAIN VARIATION OF BATAN FUEL PIN PROTOTYPE DURING IRRADIATION TEST IN RSG-GAS REACTOR*. The first short fuel pin containing natural UO2 pellet in Zry4 cladding has been prepared at the CNFT (Center for Nuclear Fuel Technology then a ramp test will be performed. The present work is part of designing first irradiation experiments in the PRTF (Power Ramp Test Facility of RSG-GAS 30 MW reactor. The thermal mechanic of the pin during irradiation has simulated. The geometry variation of pellet and cladding is modeled by taking into account different phenomena such as thermal expansion, densification, swelling by fission product, thermal creep and radiation growth. The cladding variation is modeled by thermal expansion, thermal and irradiation creeps. The material properties are modeled by MATPRO and standard numerical parameter of TRANSURANUS code. Results of irradiation simulation with 9 kW/m LHR indicates that pellet-clad contacts onset from 0.090 mm initial gaps after 806 d, when pellet radius expansion attain 0.015 mm while inner cladding creep-down 0.075 mm. A newer computation data show that the maximum measured LHR of n-UO2 pin in the PRTF 12.4 kW/m. The next simulation will be done with a higher LHR, up to ~ 25 kW/m. MODEL SIMULASI VARIASI GEOMETRI DAN STRESS-STRAIN DARI PROTOTIP BAHAN BAKAR PIN BATAN SELAMA UJI IRADIASI DI REAKTOR RSG-GAS. Pusat Teknologi Bahan Bakar Nuklir (PTBBN telah menyiapkan tangkai (pin bahan bakar pendek perdana yang berisi pelet UO2 alam dalam kelongsong paduan zircaloy untuk dilakukan uji iradiasi daya naik. Penelitian ini merupakan bagian dari perancangan percobaan iradiasi pertama di PRTF (Power Ramp Test Fasility yang terpasang di reaktor serbaguna RSG-GAS berdaya 30 MW. Telah dilakukan pemodelan dan simulasi kinerja termal mekanikal pin selama iradiasi. Variasi geometri pelet dan kelongsong selama pengujian dimodelkan dengan memperhatikan fenomena ekspansi termal

  15. Data about operation and utilization of the RA in 1976 - Report, Annex I; Prilog I - Podaci o radu i iskoriscenosti reaktora RA u 1976. godini - Izvestaj -

    Energy Technology Data Exchange (ETDEWEB)

    Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    During 1976 there were significant discrepancies in RA reactor operation in comparison with the operating plan made at the end of 1975. Some discrepancies were planned but not all could be foreseen due to the following reasons: no possibility to estimate the precise schedule of aluminium tubes for fuel channels delivery from the; USSR; no means to estimate precise time for repairs that were never performed or were not done during previous 6 years; inability to estimate the time needed for experiments within the program of transfer to highly enriched fuel (which was successfully finished) since some of these experiments were never done before. Some repairs were not planned since unpredicted failures occurred. Six fuel exchanges were planned, but they were somewhat delayed. Detailed data about the planned and achieved reactor operation are shown in this annex as well as interruptions. There were 12 shorter interruptions and safety shutdowns which was in total for 64 hours. Reactor was operated at nominal power of 6.5 MW. There have been no significant accidents, but two cases could be characterised as accidental according to their occurrence not according to the effects. The first was break of the tube next to the heavy water pump, and the second was contamination of the reactor hall and corridor during preparation of the fuel exchange machine (spill of heavy water). This annex covers the data concerned with utilization of experimental facilities and reactor utilization. [Serbo-Croat] Reaktor RA je u toku 1976. godine radio sa znatnijim odstupanjima od plana rada za 1976. godinu, donetom krajem 1975. godine. Medjutim, ni ova procena mogucih odstupanja nije mogla biti korektna iz sledecih, razloga: nemogucnost da se predvidi tacan rok isporuke cevi od specijalne aluminijumske legure iz SSSR za izradu tehnoloskih kanala, nemogucnost da se predvidi tacno vreme potrebno za remonte kakvi ili nisu radjeni ranije ili su radjeni pre vise od 6 godina pod povoljnijim uslovima

  16. Adaptive Neural Network Algorithm for Power Control in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Husam Fayiz, Al Masri

    2017-01-01

    The aim of this paper is to design, test and evaluate a prototype of an adaptive neural network algorithm for the power controlling system of a nuclear power plant. The task of power control in nuclear reactors is one of the fundamental tasks in this field. Therefore, researches are constantly conducted to ameliorate the power reactor control process. Currently, in the Department of Automation in the National Research Nuclear University (NRNU) MEPhI, numerous studies are utilizing various methodologies of artificial intelligence (expert systems, neural networks, fuzzy systems and genetic algorithms) to enhance the performance, safety, efficiency and reliability of nuclear power plants. In particular, a study of an adaptive artificial intelligent power regulator in the control systems of nuclear power reactors is being undertaken to enhance performance and to minimize the output error of the Automatic Power Controller (APC) on the grounds of a multifunctional computer analyzer (simulator) of the Water-Water Energetic Reactor known as Vodo-Vodyanoi Energetichesky Reaktor (VVER) in Russian. In this paper, a block diagram of an adaptive reactor power controller was built on the basis of an intelligent control algorithm. When implementing intelligent neural network principles, it is possible to improve the quality and dynamic of any control system in accordance with the principles of adaptive control. It is common knowledge that an adaptive control system permits adjusting the controller’s parameters according to the transitions in the characteristics of the control object or external disturbances. In this project, it is demonstrated that the propitious options for an automatic power controller in nuclear power plants is a control system constructed on intelligent neural network algorithms. (paper)

  17. Improved lumped models for transient combined convective and radiative cooling of a two-layer spherical fuel element

    International Nuclear Information System (INIS)

    Silva, Alice Cunha da; Su, Jian

    2013-01-01

    The High Temperature Gas cooled Reactor (HTGR) is a fourth generation thermal nuclear reactor, graphite-moderated and helium cooled. The HTGRs have important characteristics making essential the study of these reactors, as well as its fuel element. Examples of these are: high thermal efficiency,low operating costs and construction, passive safety attributes that allow implication of the respective plants. The Pebble Bed Modular Reactor (PBMR) is a HTGR with spherical fuel elements that named the reactor. This fuel element is composed by a particulate region with spherical inclusions, the fuel UO2 particles, dispersed in a graphite matrix and a convective heat transfer by Helium happens on the outer surface of the fuel element. In this work, the transient heat conduction in a spherical fuel element of a pebble-bed high temperature reactor was studied in a transient situation of combined convective and radiative cooling. Improved lumped parameter model was developed for the transient heat conduction in the two-layer composite sphere subjected to combined convective and radiative cooling. The improved lumped model was obtained through two-point Hermite approximations for integrals. Transient combined convective and radiative cooling of the two-layer spherical fuel element was analyzed to illustrate the applicability of the proposed lumped model, with respect to die rent values of the Biot number, the radiation-conduction parameter, the dimensionless thermal contact resistance, the dimensionless inner diameter and coating thickness, and the dimensionless thermal conductivity. It was shown by comparison with numerical solution of the original distributed parameter model that the improved lumped model, with H2,1/H1,1/H0,0 approximation yielded significant improvement of average temperature prediction over the classical lumped model. (author)

  18. The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

    Energy Technology Data Exchange (ETDEWEB)

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest

  19. KONVERSI KATALITIK MINYAK SAWIT UNTUK MENGHASILKAN BIOFUEL MENGGUNAKAN SILIKA ALUMINA DAN HZSM-5 SINTESIS

    Directory of Open Access Journals (Sweden)

    Nurjannah Nurjannah

    2012-02-01

    Full Text Available Terbatasnya sumber energi fosil menyebabkan perlunya pengembangan energi terbarukan yang berasal dari alam dan dapat diperbaharui. Penggunaan bahan bakar minyak bumi, baik dari penggunaan berupa alat transportasi maupun dari penggunaan oleh industri sangat mencemari lingkungan karena tingkat polusi yang ditimbulkan sangat tinggi sehingga perlu mencari bahan bakar alternatif pengganti bahan bakar gasoline, solar, dan kerosene dari minyak nabati. Penelitian dilakukan dalam dua tahapan yaitu sintesa katalis dan proses katalitik cracking. Silika alumina disintesa menggunakan metode Latourette dan HZSM-5 disintesa menggunakan metode Plank. Hasil sintesa dikarakterisasi dengan Penyerapan Spektroskopi Atomis (AAS menunjukkan bahwa silika alumina dan HZSM-5 mempunyai Si/Al 198 dan 243. Luas permukaan  silika alumina dan HZSM-5 diperoleh dari analisa Brunauer Emmet Teller (BET yaitu 149,91-213,35 m2.g-1 dan ukuran pori rata-rata adalah 13oA. Perengkahan katalitik dilakukan dalam suatu mikroreaktor fixed bed pada temperatur 350-500°C dan laju alir gas N2 100-160 ml.min-1 selama 120 min. Hasil perengkahan dianalisa dengan metode gas kromatografi. Hasil yang diperoleh untuk katalis HZSM-5 fraksi gasoline dengan yield tertinggi 28,87%, kerosene 16,70%, dan diesel 12,20%  pada suhu reaktor 4500C dan laju gas N2 100 ml/menit.

  20. Operational experience gained with the failed fuel rod detection system in nuclear power plants

    International Nuclear Information System (INIS)

    Boehm, H.H.; Forch, H.

    1985-01-01

    Brown Boveri Reaktor GmbH together with Krautkramer Company developed such a FAILED FUEL ROD DETECTION SYSTEM (FFRDS) which allows to located defective fuel rods without dismantling the fuel assembly or pulling of individual rods. Since 1979 the FFRDS is employed successfully in various nuclear power plants in Europe, USA, Japan, and Korea. The short inspection time and the high reliability of the method make the FFRDS a true competitor to the sipping method. In this paper the authors discuss the method and the design of the system, the equipment set-up, its features and the experience gained so far. The system has been performed and automated to such an extent that within a short installation period series of fuel assemblies can be tested with relatively short intervals of time (5 minutes for BWR and 7 minutes for PWR fuel assemblies per side). The ability of the system for deployment under various conditions and the experience gained during the past six years have made this system universally applicable and highly sensitive to the requirements of NDT during outages and for transport of FAs to intermediate storage facilities. Comparison of FFRDS to conventional sipping has indicated in several instances that the FFRDS is superior to the latter technique

  1. Combination of biological treatment of waste air and liquid effluents in small and medium-sized businesses by using spherical carrier media in exchange with each other. Final report; Kombination der biologischen Behandlung von Abluft- und Abwasserreinigung in kleinen und mittleren Unternehmen durch die Verwendung von kugeligen Traegermaterialien im gegenseitigen Austausch. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kunz, P.M.; Brunk, M.; Bentz, P.; Bach, K.; Marcos del Rio, O.; Stahl, B.; Wolf, C.

    2002-05-08

    A combined system for biological treatment of solvent-contaminated liquid and gaseous effluents was to be constructed. The system was to be mobile for demonstration purposes in industrial processes and was to serve for dimensioning of full-scale systems. The system works by a combined process via exchange of spherical carriers with a growth of micro-organisms. Two pilot systems of the Institut fuer Biologische Verfahrenstechnik, one for liquid and one for gaseous effluents, were combined for this purpose, ensuring exchange of carrier materials between the two units. (orig.) [German] Aufgabe im Rahmen dieses F+E-Vorhabens war es gewesen, eine kombinierte biologische Abluft- und Abwasserreinigungsanlage - speziell fuer loesungsmittelhaltige Abluft und Abwaesser - zu untersuchen. Dazu sollte eine Anlage aufgebaut werden, die auch als Demonstrations- und Versuchsanlage zu Industriebetrieben gebracht und dort zur Dimensionierung von full-scale-Anlage eingesetzt werden kann. Idee fuer dieses Vorhaben war gewesen, die Kombination beider Behandlungssysteme ueber den Austausch von mit Mikroorganismen bewachsenem Traegermaterial herzustellen. Am Institut fuer Biologische Verfahrenstechnik existierte bereits eine Pilotanlage fuer die biologische Abwasserreinigung mit kugeligem Traegermaterial (Vario-Bio-Reaktor) und eine Pilotanlage fuer die biologische Behandlung von Abluft nach dem Trickling-Filter-Prinzip mit ebenfalls kugeligem Traegermaterial. Diese beiden Anlagen sollten miteinander kombiniert und der Austausch des Traegermaterials hergestellt werden. (orig.)

  2. Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

    Science.gov (United States)

    Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

    The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

  3. Failed fuel rod detection system and computerized manipulator during outages

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1984-01-01

    During regular outages spent fuel assemblies need to be replaced and relocated within the core. Defective fuel rods in particular fuel assemblies have to be removed from further service and before delivery of such faulty fuel assemblies to a reprocessing plant. The system which Brown Boveri Reaktor GmbH and Krautkraemer have developed in the Federal Republic of Germany is capable of directly locating the defective rods in a proper fuel assembly. Inspection times are comparable to those of standard sipping methods, with the advantages of immediately available results and direct identification of the defective fuel rods. During the repair of fuel assemblies this system allows withdrawal of individual defective rods. With the sipping method all the fuel rods of a defective fuel assembly need to be removed and inspected by eddy current testing. During steam generator inspection and repair personnel are exposed to ample radiation. A remotely controlled, computerized manipulator was used to significantly reduce the radiation dose by automating steps in the procedures; at the same time inspection and repair times were reduced. The main features of the manipulator are a rigid component construction of the leg and two arms, and a resolver control for horizontal and vertical motion that enables rapid and accurate access to a desired tube (author)

  4. Introducing nuclear power into currently non-nuclear states

    International Nuclear Information System (INIS)

    Gert, Claassen

    2007-01-01

    As the nuclear renaissance gains momentum, many countries that currently have no nuclear power plants will begin to consider introducing them. It is anticipated that smaller reactors such as the Pebble Bed Modular Reactor (PBMR) will not only be sold to current nuclear states to also to states where there is currently no nuclear experience. A range of issues would have to be considered for nuclear plants to be sold to non-nuclear states, such as the appropriate regulatory environment, standardization and codes, non-proliferation, security of supply, obtaining experienced merchant operators, appropriate financial structures and education and training. The paper considers nine major issues that need to be addressed by governments and vendors alike: 1) political enabling framework, 2) regulatory framework, 3) responsible owner, 4) responsible operator, 5) finance, 6) contact management, 7) fuel supply and waste management framework, 8) training and education, and 9) industrial infrastructure. International cooperation by organisations such as the IAEA, financial institutions and international suppliers will be required to ensure that developing countries as well as developed ones share the benefits of the nuclear renaissance. The opportunities that the nuclear industry affords to develop local skills, create job opportunities and to develop local manufacturing industries are among the important reasons that the South African Government has decided to support and fund the development of the Pebble Bed Modular Reactor project. (author)

  5. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  6. Basal Metabolic Rate of Adolescent Modern Pentathlon Athletes: Agreement between Indirect Calorimetry and Predictive Equations and the Correlation with Body Parameters

    Science.gov (United States)

    Loureiro, Luiz Lannes; Fonseca, Sidnei; Castro, Natalia Gomes Casanova de Oliveira e; dos Passos, Renata Baratta; Porto, Cristiana Pedrosa Melo; Pierucci, Anna Paola Trindade Rocha

    2015-01-01

    Purpose The accurate estimative of energy needs is crucial for an optimal physical performance among athletes and the basal metabolic rate (BMR) equations often are not well adjusted for adolescent athletes requiring the use of specific methods, such as the golden standard indirect calorimetry (IC). Therefore, we had the aim to analyse the agreement between the BMR of adolescents pentathletes measured by IC and estimated by commonly used predictive equations. Methods Twenty-eight athletes (17 males and 11 females) were evaluated for BMR, using IC and the predictive equations Harris and Benedict (HB), Cunningham (CUN), Henry and Rees (HR) and FAO/WHO/UNU (FAO). Body composition was obtained using DXA and sexual maturity data were retrieved through validated questionnaires. The correlations among anthropometric variables an IC were analysed by T-student test and ICC, while the agreement between IC and the predictive equations was analysed according to Bland and Altman and by survival-agreement plotting. Results The whole sample average BMR measured by IC was significantly different from the estimated by FAO (pBMR when compared with IC (T Test). When compared to the golden standard IC, using Bland and Altman, ICC and Survival-Agreement, the equations underestimated the energy needs of adolescent pentathlon athletes up to 300kcal/day. Therefore, they should be used with caution when estimating individual energy requirements in such populations. PMID:26569101

  7. Anaerobic degradation of tetrachloroethylene; Anaerober Abbau von Tetrachlorethylen

    Energy Technology Data Exchange (ETDEWEB)

    Diekert, G. [Stuttgart Univ. (Germany). Inst. fuer Mikrobiologie; Scholz-Muramatsu, H. [Stuttgart Univ. (Germany). Inst. fuer Siedlungswasserbau

    1996-12-31

    Dehalospirillum multivorans, a tetrachloroethylene-dechlorinating bacterium, was isolated in activated sludge. This organism is able to grow on a defined medium with hydrogen and tetrachloroethylene (PCE) as its only energy source. The organism was characterised and the physiology of dechlorination was studied. In this process PCE is dechlorinated to cis-1,2-dichloroethene (DCE) via trichloroethene (TCE). A fluidized-bed reactor which reduces PCE to DCE at a high rate (15 nmol/min/mg of protein at 5 {mu}M PCE) was inoculated with the bacterium. Meanwhile a reactor inoculated with D. multivorans and a fully dechlorinating mixed culture has become available which catalyses the complete dechlorination of PCE to ethene at just as high rates. Tetrachloroethene dehalogenase was purified from D. multivorans (unpublished results) and characterised. (orig./SR) [Deutsch] Aus Belebtschlamm wurde ein Tetrachlorethen-dechlorierendes Bakterium, Dehalospirillum multivorans, isoliert. Der Organismus waechst auf definiertem Medium mit Wasserstoff und Tetrachlorethen (PCE) als einziger Energiequelle. Der Organismus wurde charakterisiert und die Physiologie der Dechlorierung wurde untersucht. PCE wird dabei ueber Trichlorethen (TCE) bis zum cis-1,2-Dichlorethen (DCE) dechloriert. Mit diesem Bakterium wurde ein Wirbelschichtreaktor inokuliert, der mit hohen Raten (15 nmol/min/mg Protein bei 5 {mu}M PCE) PCE zu DCE reduziert. Inzwischen steht ein Reaktor zur Verfuegung, der mit D. multivorans und einer voellig dechlorierenden Mischkultur inokuliert wurde und der mit ebenso hohen Raten eine vollstaendige Dechlorierung von PCE bis zum Ethen katalysiert. Aus D. multivorans wurde die Tetrachlorethen-Dehalogenase gereinigt (unveroeffentlichte Ergebnisse) und charakterisiert. (orig./SR)

  8. Anaerobic degradation of tetrachloroethylene; Anaerober Abbau von Tetrachlorethylen

    Energy Technology Data Exchange (ETDEWEB)

    Diekert, G [Stuttgart Univ. (Germany). Inst. fuer Mikrobiologie; Scholz-Muramatsu, H [Stuttgart Univ. (Germany). Inst. fuer Siedlungswasserbau

    1997-12-31

    Dehalospirillum multivorans, a tetrachloroethylene-dechlorinating bacterium, was isolated in activated sludge. This organism is able to grow on a defined medium with hydrogen and tetrachloroethylene (PCE) as its only energy source. The organism was characterised and the physiology of dechlorination was studied. In this process PCE is dechlorinated to cis-1,2-dichloroethene (DCE) via trichloroethene (TCE). A fluidized-bed reactor which reduces PCE to DCE at a high rate (15 nmol/min/mg of protein at 5 {mu}M PCE) was inoculated with the bacterium. Meanwhile a reactor inoculated with D. multivorans and a fully dechlorinating mixed culture has become available which catalyses the complete dechlorination of PCE to ethene at just as high rates. Tetrachloroethene dehalogenase was purified from D. multivorans (unpublished results) and characterised. (orig./SR) [Deutsch] Aus Belebtschlamm wurde ein Tetrachlorethen-dechlorierendes Bakterium, Dehalospirillum multivorans, isoliert. Der Organismus waechst auf definiertem Medium mit Wasserstoff und Tetrachlorethen (PCE) als einziger Energiequelle. Der Organismus wurde charakterisiert und die Physiologie der Dechlorierung wurde untersucht. PCE wird dabei ueber Trichlorethen (TCE) bis zum cis-1,2-Dichlorethen (DCE) dechloriert. Mit diesem Bakterium wurde ein Wirbelschichtreaktor inokuliert, der mit hohen Raten (15 nmol/min/mg Protein bei 5 {mu}M PCE) PCE zu DCE reduziert. Inzwischen steht ein Reaktor zur Verfuegung, der mit D. multivorans und einer voellig dechlorierenden Mischkultur inokuliert wurde und der mit ebenso hohen Raten eine vollstaendige Dechlorierung von PCE bis zum Ethen katalysiert. Aus D. multivorans wurde die Tetrachlorethen-Dehalogenase gereinigt (unveroeffentlichte Ergebnisse) und charakterisiert. (orig./SR)

  9. Change of I-V characteristics of SiC diodes upon reactor irradiation; Modification des caracteristiques I-V de jonctions p-n au SiC du fait d'une irradiation dans un reacteur; Izmeneniya kharakteristik I-V vyrashchennogo v SiC perekhoda tipa p-n posle oblucheniya ego v reaktore; Modificaciones que sufren por irradiacion en un reactor las caracteristicas I-V de uniones p-n en SiC

    Energy Technology Data Exchange (ETDEWEB)

    Heerschap, M; De Coninck, R [Solid State Physics Dept., SCK-CEN, Mol (Belgium)

    1962-04-15

    distintas procedencias. Las uniones se obtuvieron en el horno de Lely. Midieron las caracteristicas directa e inversa durante la irradiacion, y despues de esta, hasta la temperatura de 150{sup o}C. Se estan realizando mediciones hasta 500{sup o}C. Han encontrado que uno de los tipos de diodo es resistente a los neutrones del BR-1 hasta un flujo integrado de 10{sup 15} neutrones/cm{sup 2}, mientras que el otro soporta hasta 10{sup 17} neutrones/cm{sup 2}. La memoria indica los cambios de las caracteristicas, asi como los resultados de algunos experimentos de recocido. (author) [Russian] V poiskakh poluprovodnikov, kotorye mogli by byt' ispol'zovany v reaktorakh s vysokoj plotnost'yu nejtronnogo potoka dlya izmereniya raspredeleniya potokov, my obluchali v bel'gijskom reaktore BR-1 perekhody tipa p-n v SiC. Byli oblucheny dva tipa diodov SiC razlichnogo proiskhozhdeniya. EHti perekhody vyrashchivayutsya v pechi Loli. Izmeneniya pryamoj i obratnoj kharakteristik byli izmereny posle i vo vremya oblucheniya vplot' do temperatury 150{sup o}C; v nastoyashchee vremya proizvodyatsya izmereniya vplot' do temperatury 500{sup o}C. Bylo ustanovleno, chto odin tip dioda vyderzhivaet nejtronnoe obluchenie reaktora BR-1 vplot' do integrirovannogo potoka 10{sup 15} nejtronov na kv. sm, togda kak drugoj tip vyderzhivaet obluchenie vplot' do potoka 10{sup 17} nejtronov na kv. sm. Dayutsya izmeneniya kharakteristik, a takzhe rezul'taty nekotorykh ehksperimentov otzhiga. (author)

  10. Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

    2012-01-01

    Highlights: ► Models for PBMR and thermochemical sulfur cycle based hydrogen plant are developed. ► Models are validated against available data in literature. ► Transient in coupled reactor and hydrogen plant system is studied. ► For loss-of-heat sink accident, temperature feedback within the reactor core enables shut down of the reactor. - Abstract: A high temperature reactor (HTR) is a candidate to drive high temperature water-splitting using process heat. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Three thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. The models and coupling scheme are presented here, as well as a transient test case initiated within the chemical plant. The 50% feed flow failure within the chemical plant results in a slow loss-of-heat sink (LOHS) accident in the nuclear reactor. Due to the temperature feedback within the reactor core the nuclear reactor partially shuts down over 1500 s. Two distinct regions are identified within the coupled plant response: (1) immediate LOHS due to the loss of the sulfuric

  11. A comprehensive economic evaluation of integrated desalination systems using fossil fuelled and nuclear energies and including their environmental costs

    International Nuclear Information System (INIS)

    Nisan, S.; Benzarti, N.

    2008-01-01

    Seawater desalination is now widely accepted as an attractive alternative source of freshwater for domestic and industrial uses. Despite the considerable progress made in the relevant technologies desalination, however, remains an energy intensive process in which the energy cost is the paramount factor. This Study is a first of a kind in that we have integrated the environmental costs into the power and desalination costs. The study has focused on the seawater desalination cost evaluation of the following systems. It is supposed that they will be operating in the co-generation mode (Simultaneous production of electrical power and desalted water) in 2015: Fossil fuelled based systems such as the coal and oil fired plants and the gas turbine combined cycle plant, coupled to MED, and RO; Pressurised water reactors such as the PWR-900 and the AP-600, coupled to MED, and RO; High temperature reactors such as the GT-MHR, the PBMR, coupled to MED, with the utilisation of virtually free waste-heat provided by these reactors. The study is made in real site-specific conditions of a site In Southern Europe. Sensitivity studies for different parameters such as the fossil fuel prices, interest and discount rates, power costs etc., have also been undertaken. The results obtained are then used to evaluate the financial interest of selected integrated desalination systems in terms of a detailed cash flow analysis, providing the net present values, pay back periods and the internal rate of returns. Analysis of the results shows that among the fossil fuelled systems the power and desalination costs by circulating fluidized bed coal fired plant would be the lowest with current coal prices. Those by oil fired plants would be highest. In all cases, integrated nuclear energy systems would lead to considerably lower power and water costs than the corresponding coal based systems. When external costs for different energies are internalized in power and water costs, the relative cost

  12. Ways to increase efficiency of the HTGR coupled with the gas-turbine power conversion unit - HTR2008-58274

    International Nuclear Information System (INIS)

    Golovko, V. F.; Kodochigov, N. G.; Vasyaev, A. V.; Shenoy, A.; Baxi, C. B.

    2008-01-01

    The paper deals with the issue of increasing efficiency of nuclear power plants with the modular high-temperature helium reactor (HTGR) and direct gas turbine cycle. It should be noted that only this combination can highlight the advantages of the HTGR, namely the ability to heat helium to about 1000 deg. C, in comparison with other reactor plants for electricity generation. The HTGR has never been used in the direct gas turbine cycle. At present, several designs of such commercial plants are at the stage of experimental validation of main technical features. In Russia, 'OKB Mechanical Engineering' together with 'General Atomics' (USA) are developing the GT-MHR project with the reactor power of 600 MW, reactor outlet helium temperature of 850 deg. C, and efficiency of about 45.2%; the South African Republic is developing the PBMR project with the reactor power of 400 MW, reactor outlet helium temperature of 900 deg. C, and efficiency of about 42%; and Japan is developing the GTHTR-300 project with the reactor power of 600 MW, reactor outlet helium temperature of 850 deg. C, and efficiency of about 45.6%. As it has been proven by technical and economic estimations, one of the most important factors for successful promotion of reactor designs is their net efficiency, which must be not lower than 47%. A significant advantage of a reactor plant with the HTGR and gas-turbine power conversion unit over the steam cycle is considerable simplification of the power unit layout and reduction of the required equipment and systems (no steam generators, no turbine hall including steam lines, condenser, deaerator, etc.), which makes the gas-turbine power conversion unit more compact and less costly in production, operation and maintenance. However, in spite of this advantage, it seems that in the projects currently being developed, the potential of the gas-turbine cycle and high-temperature reactor to more efficiently generate electricity is not fully used. For example, in modern

  13. Research nuclear reactor RA - Annual Report 1975. Operation and maintenance; Istrazivacki nuklearni reaktor RA - Izvestaj za 1975. godinu - Pogon i odrzavanje

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-01-15

    The plan for 1975 was successfully fulfilled. This is reflected in research related to improvement of operating properties of the RA reactor, mostly due to the effort of the RA staff employed in operation and maintenance of the reactor. Fuel saving achieved by this activity amounted to about 38% (80% enriched fuel). Preliminary work is done, concerned with new reactor core with highly enriched fuel. This is a significant saving as well. New fuel elements have arrived at the end of this year. It is going to enable increase of neutron flux by 50% without changing the nominal operating power. The possibility of further improvement of the reactor are analyzed, to enable material testing and production of radioactive sources. Mid term plan for reactor operation was made according to this analysis. It is planned to further increase the neutron flux in isolated smaller zones, and building new experimental loops with cooling and fast neutron converters. Much was done to increase the safety level of reactor operation and preparing the safety report. [Serbo-Croat] Izvrsenje zadataka u 1975. godini bilo je uspesno. To se ogleda u povecanju istrazivackog rada vezanog za poboljsanje eksploatacionih karakteristika reaktora RA, pretezno koriscenjem sopstvenog kadra angazovanog u pogonu i odrzavanju reaktora. Ovim radom postignuta je usteda goriva od oko 38% (80% obogaceno gorivo). Izvrseni su preliminarni radovi na prevodjenju reaktora RA na novo gorivo, sto je takodje velika usteda. Novo gorivo je stiglo krajem godine i ono ce obezbediti porast neutronskog fluksa od 50%, bez promene nominalne snage reaktora. Izvrsena je analiza mogucnosti daljeg usavrsavanja reaktora za potrebe ispitivanja materijala kao i proizvodnju radioaktivnih izvora. Na osnovu ove analize nacinjen je srednjorocni program rada reaktora RA sa tezistem na daljem povecanju fluksa u izdvojenim manjim zonama i ugradnju 'hladjenih petlji' i brzih konvertora. Mnogo je ucinjeno na povecanju stepena sigurnosti reaktora RA i izradi sigurnosnog izvestaja.

  14. Pengaruh Aerasi dan Sumber Nutrien terhadap Kemampuan Alga Filum Chlorophyta dalam Menyerap Karbon (Carbon Sink untuk Mengurangi Emisi CO2 di Kawasan Perkotaan

    Directory of Open Access Journals (Sweden)

    Lancur Setoaji

    2013-09-01

    Full Text Available Penelitian terkait mitigasi pemanasan global, khususnya dalam penyerapan karbon dioksida (CO2, menjadi fokus utama di kalangan ilmuwan dunia. Secara alamiah, karbon dioksida dapat diserap oleh tumbuhan hijau, laut, karbonasi batuan kapur, dan alga. Pigmen hijau dalam alga atau klorofil dapat menyerap karbon dioksida dalam proses fotosintesis. Alga memiliki pertumbuhan yang sangat cepat sehingga cocok digunakan sebagai carbon sink. Penelitian terkait carbon sink ini bertujuan untuk menentukan kemampuan rata-rata serapan CO2 oleh alga di kawasan perkotaan dan menentukan pengaruh aerasi dan variasi sumber N terhadap pertumbuhan dan perkembangan alga. Penelitian ini dilakukan dalam skala laboratorium menggunakan reaktor dengan proses batch. Sampel alga yang digunakan didapatkan dari hasil pengembangbiakan yang bersumber dari perairan di kawasan perkotaan. Penelitian ini menggunakan dua variabel uji, yaitu aerasi dan sumber nutrien. Jumlah karbon dioksida yang diserap didapatkan dari perbandingan stoikiometri pada reaksi fotosintesis.  Berdasarkan perbandingan stoikiometri tersebut diketahui bahwa 1 gram sel alga yang terbentuk sebanding dengan 1,92 gram CO2 yang diserap. Dari hasil penelitian, alga dengan penambahan pupuk urea dapat menyerap 4,87 mg CO2/hari dalam kondisi tanpa aerasi atau 3,84 mg CO2/hari dengan aerasi. Sedangkan alga dengan penambahan pupuk NPK dapat menyerap 3,61 mg CO2/hari dalam kondisi tanpa aerasi atau 3,01 mg CO2/hari dengan aerasi.

  15. Operation, maintenance and utilization of the RA reactor, Annual report for 1980; Pogon, odrzavanje i eksploatacija reaktora RA, Izvestaj o radu u 1980. godini

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1980-12-15

    During 1980 the activities of RA reactor staff was conducted in two directions: repair works and determination of the state of the existing equipment; and preparing and constructing additional equipment and completing the documentation needed to fulfill the new legal regulations. preparation of the existing equipment for future operation was finished, but construction of the additional equipment (according to the regulations) is not done at planned rate due to different reasons such as lack of budget, administration, import. This report includes a chapter devoted to the analysis of the operation difficulties that caused prohibition of reactor operation as well as difficulties in achievement of the planned activities for 1980. Data about financial issues are included as well. [Serbo-Croat] U toku 1980. godine rad OOUR-a Nuklearni reaktor RA odvijao se u dva smera: remontnim radovima i utvrdjivanju stanja postojece opreme i u pripremi i ugradnji potrebne dodatne opreme i kompletiranju neophodne dokumentacije prema novim zakonskim propisima. Priprema postojece opreme za rad je u potpunosti izvrsena, dok se projekti i ugradnja dodatne opreme (prema zakonskim obavezama) ne odvija potrebnom brzinom zbog raznih objektivnih razloga, kao nedostatak sredstava, odobrenja, uvoza. Ovaj izvestaj sadrzi poglavlje posveceno analizi teskoca u radu koje su dovele do zabrane rada reaktora, i teskoca u realizovanju plana rada u 1980. godini, ako i podatke o finansiranju projekta 'pogon, odrzavanje i eksploatacija reaktora RA'.

  16. Investigation of corrosion of materials of the irradiation device in the RA reactor; Ispitivanje korozije materijala uredjaja za ozracivanje na reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Zaric, M; Mance, A; Vlajic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    Devices for sample irradiation in the vertical RA reactor channels will be made of aluminium alloys. According to the regulations concerned with introducing materials into the RA reactor core, corrosion characterisation of these materials is an obligation. Corrosion properties of four aluminium alloys were investigated both in contact with stainless steel and without it. First part of this report deals with the corrosion testing of aluminium alloys in water by gravimetric and electrochemical methods. Bi-distilled water at temperatures less than 100 deg C was used. Second part is related to aluminium alloys corrosion in carbon dioxide gas under experimental conditions. The second part of research was initiated by the design of the head of the independent CO{sub 2} loop for samples cooling. [Serbo-Croat] Uredjaji za ozracivanje u vertikalnim kanalima reaktora RA, bice napravljeni od legura aluminjuma. Prema propisima o unosenju materijala u RA reaktor materijali se moraju prethodno ispitati i sa stanovista korozije. Ispitivane su korozione pojave na cetiri aluminjumske legure sa i bez kontakta sa nerdjajucim celikom. Prvi deo ovog rada tretira pitanje korozije legura aluminijuma u vodi gravimetrijskim i elektrohemijskim metodama. Koriscena je bidestilovana voda na temperaturi do 100 deg C. Drugi deo se odnosi na ispitivanje ponasanja legura aluminijuma u gasovitom ugljen dioksidu pod uslovima eksperimenta. Drugi deo istrazivanja izvrsen je za potrebe izgradnje glave petlje nezavisnog kola za hladjenje uzoraka gasovitim CO{sub 2}.

  17. Radioactivity of spent TRIGA fuel

    International Nuclear Information System (INIS)

    Usang, M. D.; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-01-01

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive

  18. Nuclear fuel in water reactors: Manufacturing technology, operational experience and development objectives in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Holzer, R.; Knoedler, D.

    1977-01-01

    The nuclear fuel industry in the Federal Republic of Germany comprises the full range of manufacturing capabilities for pressurized-, boiling- and heavy-water reactor technology. The existing manufacturing companies are Reaktor-Brennelement Union (RBU) and Alkem. RBU makes natural and enriched UO 2 -fuel assemblies, starting with powder preparation. Facilities to produce UO 2 -gadolinia and UO 2 -ThO 2 fuel are also available. Alkem manufactures mixed-oxide UO 2 /PuO 2 fuel and fuel rods. Zircaloy cladding tubes are produced by Nuklearrohr-Gesellschaft (NRG) and Mannesmannroehren-Werke (MRW). Construction of a new fuel manufacturing plant has been announced by Exxon. Supplementary to quality control, an integrated quality assurance system has been established between the reactor vendor's fuel design and engineering division and the existing manufacturing companies for fuel and tubing. Operating experience with LWR and HWR fuel dates back to 1964/65 and has shown good performance. Possible reasons for a small fraction of defective rods could be identified quickly by a fast feedback system incorporating close co-operation between Kraftwerk Union (KWU) and the utilities. KWU combines fuel development, hot-cell and pool-side service facilities as well as fuel technology linked to manufacturing. The responsibility of KWU for core and fuel design, which enabled an integral optimization, was also an important reason for the successful operation and design flexibility. (author)

  19. Radioactivity of spent TRIGA fuel

    Energy Technology Data Exchange (ETDEWEB)

    Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P. [Reactor Department, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  20. Selective separation of anaerobic sludge by means of hydrocyclones; Selektive Abtrennung von Anaerobschlamm mit Hydrozyklonen

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, M.; Bohnet, M. [Technische Univ. Braunschweig (Germany). Inst. fuer Verfahrens- und Kerntechnik

    1999-07-01

    In anaerobic waste water cleaning, biomass concentration constitutes a central problem because of long generating times and low biomass sinking speeds. In order to decouple hydraulic retention time from biomass retention time, biomass must be fed back into the reactor. The fact that separation by means of common gravitational separators such as sedimentation tanks and baffle plate thickeners is unspecific results in the enrichment in the reactor of inorganic solids, whose presence is corollary to the anaerobic sludge process. Hence, industry has a great interest in separating anaerobic sludge into organic and inorganic constituents as a means of safeguarding high operating stability and degradation efficiency of anaerobic reactors. Hydrocyclones, permitting selective separation, are an obvious approach. (orig.) [German] Bei der anaeroben Abwasserreinigung ist die Biomassekonzentrierung aufgrund langer Generationszeiten und geringer Sinkgeschwindigkeiten der Biomasse ein zentrales Problem. Zur Entkopplung der hydraulischen Verweilzeit von der Verweilzeit der Biomasse ist eine Rueckfuehrung der Biomasse erforderlich. Da bisher eingesetzte Schwerkraftabscheider, wie Absetzbecken und Lamellenklaerer, unspezifisch trennen, kommt es zu einer Anreicherung anorganischer Feststoffe im Reaktor, die sich prozessbedingt im Anaerobschlamm befinden. So hat die Industrie ein grosses Interesse an einer Auftrennung des Anaerobschlamms in organische und anorganische Bestandteile, um eine hohe Betriebsstabilitaet und Abbauleistung der Anaerobreaktoren zu gewaehrleisten. Hierzu bieten sich Hydrozyklone an, weil mit ihnen eine selektive Trennung moeglich ist. (orig.)

  1. Fluidized-bed gasification of biomass: Conversion of fine carabon particles in the freeboard; Biomassevergasung in der Wirbelschicht: Umsatz von feinen Kohlenstoffpartikeln im Freeboard

    Energy Technology Data Exchange (ETDEWEB)

    Miccio, F [Ist. Ricerche sulla Combustione-CNR, Napoli (Italy); Moersch, O; Spliethoff, H; Hein, K R.G. [Stuttgart Univ. (Germany). Inst. fuer Verfahrenstechnik und Dampfkesselwesen

    1998-09-01

    The conversion of carbon particles in gasification processes was investigated in a fluidized-bed reactor of the Institute of Chemical Engineering and Steam Boiler Technology of Stuttgart University. This reactor is heated electrically to process temperature, and freeboard coal particles can be sampled using an isokinetic probe. The fuel used in the experiments consisted of beech wood chips. The temperature and air rating, i.e. the main parameters of the process, were varied in order to investigate their influence on product gas quality and carbon conversion. The conversion rate is influenced to a significant extent by grain disintegration and discharge of carbon particles. In gasification conditions, a further conversion process takes place in the freeboard. (orig.) [Deutsch] In dieser Arbeit wird die Umsetzung von Kohlenstoffpartikeln unter Vergasungsbedingungen untersucht. Die Versuche wurden an einem Wirbelschichtreaktor des Instituts fuer Verfahrenstechnik und Dampfkesselwesen der Universitaet Stuttgart durchgefuehrt. Dieser Reaktor wird elektrisch auf Prozesstemperatur beheizt. Mit Hilfe einer isokinetischen Sonde koennen Proben von Kohlenstoffpartikeln im Freeboard genommen werden. Als Brennstoff wurden zerkleinerte Buchenholz-Hackschnitzel eingesetzt. Variiert wurden als Hauptparameter des Prozesses Temperatur und Luftzahl. Untersucht wurde der Einfluss dieser Parameter auf die Qualitaet des Produktgases und die Umsetzung des Kohlenstoffes. Kornzersetzungs- und Austragsvorgaenge von Kohlenstoffpartikeln spielen eine wichtige Rolle fuer den Kohlenstoffumsatz. Unter Vergasungsbedingungen findet im Freeboard eine weitere Umsetzung der Partikel statt. (orig.)

  2. Safety report for the Independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels, Vol. VI; Album VI: Izvestaj o sigurnosti za nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u VEK reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-07-01

    First part of the safety report for the Independent CO{sub 2} loop for cooling the samples irradiated in the RA reactor vertical experimental channels contains descriptions of the independent CO{sub 2} loop, system for regulation, measurement and control od the loop parameters, description of the dosimetry system, and the plan for testing the experimental device before start-up. Second part of this analysis describes the influence of of the experimental device on the reactor operation under steady state conditions as follows: influence of the head of the independent coolant loop on the reactivity of the reactor core and influence on the reactor temperature coefficient. Third part of the report includes the analysis of possible accidents during operation of the independent CO{sub 2} coolant loop in the reactor. Izvestaj o sigurnosti za nezavisno kolo CO{sub 2} za hladjenje uzoraka ozracivanih u VEK reaktora RA sadrzi u prvom delu: opis nezavisnog kola CO{sub 2}, sistema za regulaciju, merenje i kontrolu parametara nezavisnog kola, sistema dozimetrijske kontrole i plan ispitivanja eksperimentalnih uredjaja pre pustanja u rad. Drugi deo ove analize obuhvata uticaj eksperimentalnog uredjaja na reaktor u normalnom rezimu rada i to: uticaj glave petlje nezavisnog kola CO{sub 2} za hladjenje uzoraka na reaktivnost reaktora i uticaj uredjaja na temperaturni koeficijent reaktora. Treci deo sadrzi analizu mogucih akcidenata u toku rada nezavisnog kola CO{sub 2} u reaktoru.

  3. SINTESA GULA DARI SAMPAH ORGANIK DENGAN PROSES HIDROLISIS MENGGUNAKAN KATALIS ASAM

    Directory of Open Access Journals (Sweden)

    Deddy Irawan

    2012-11-01

    Full Text Available SYNTHESIS OF SUGAR FROM ORGANIC WASTES VIA ACID CATALYSTHYDROLYSIS. Hydrolysis process is an important step from every process to produce biofuel withorganic wastes as raw material. Hydrolysis process with chemical uses hydrochloride acid as catalystin which will transform holocellulose to glucose. Raw material of 100 grams is put into hydrolysisreactor with batch system equipped with pressure control and ratio hydrochloride of 1 : 6 w/v. Thevariables studied were temperature of 110-140oC, HCl concentration of 0.5-1%, time of hydrolysis of15-60 minutes. The sugar concentration was taken and then be analyzed by Nelson-Somogy method.The hydrolysis, which was carried out with the temperature of 120oC, time of 30 minutes, HClconcentration of 0.75%, and the pressure of 6 bar, produced sugar reduction of 27.3 mg/mL and yieldof 15.07%. Proses hidrolisis merupakan satu tahap penting dari rangkaian tahapan proses produksi bahan bakarnabati menggunakan bahan baku sampah organik. Proses hidrolisis secara kimiawi menggunakanHCl sebagai katalis akan mengubah holoselulosa yang terdapat pada sampah organik menjadi gula.Gula yang dihasilkan inilah yang dapat difermentasi menjadi bahan bakar nabati. Bahanbaku sebanyak 100 g dimasukkan dalam reaktor hidrolisis sistem batch yang dilengkapi denganpengukur tekanan dan ditambahkan larutan HCl pada perbandingan 1:6 b/v. Hidrolisis dilakukandengan memvariasikan suhu operasi 100-140oC, waktu proses 15-60 menit, serta konsentrasi HCl 0,5-1%. Hidrolisat yang dihasilkan dianalisis kadar gula menggunakan metode Nelson-somogy. Hasilhidrolisis yang dilakukan pada suhu 120oC selama 30 menit serta konsentrasi HCl 0,75% dan tekananterukur 6 bar menghasilkan gula 27,30 mg/mL dan yield gula sebesar 15,07%.

  4. The Nuclear Renaissance — Implications on Quantitative Nondestructive Evaluations

    Science.gov (United States)

    Matzie, Regis A.

    2007-03-01

    The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not even have access to electricity. With this increased energy demand and the high and volatile price of fossil fuels, nuclear energy is experiencing resurgence. This so-called nuclear renaissance is broad based, reaching across Asia, the United States, Europe, as well as selected countries in Africa and South America. Some countries, such as Italy, that have actually turned away from nuclear energy are reconsidering the advisability of this design. This renaissance provides the opportunity to deploy more advanced reactor designs that are operating today, with improved safety, economy, and operations. In this keynote address, I will briefly present three such advanced reactor designs in whose development Westinghouse is participating. These designs include the advanced passive PWR, AP1000, which recently received design certification for the US Nuclear Regulatory Commission; the Pebble Bed Modular reactor (PBMR) which is being demonstrated in South Africa; and the International Reactor Innovative and Secure (IRIS), which was showcased in the US Department of Energy's recently announced Global Nuclear Energy Partnership (GNEP), program. The salient features of these designs that impact future requirements on quantitative nondestructive evaluations will be discussed. Such features as reactor vessel materials, operating temperature regimes, and new geometric configurations will be described, and mention will be made of the impact on quantitative nondestructive evaluation (NDE) approaches.

  5. Use of virtual environments to reduce the construction costs of the next generation nuclear power reactors

    International Nuclear Information System (INIS)

    Whisker, V.E.; Baratta, A.J.

    2007-01-01

    The near term deployment of the next generation of reactors will only be successful if they are built on time and without the costly overruns experienced in the previous generation. One critical factor in achieving these goals is to ensure the design is optimized for constructability. In this work the authors explored the effectiveness of full-scale virtual reality simulation in the optimization of the design and construction of the next generation of nuclear reactors. The research tested the suitability of immersive virtual reality display technology in aiding engineers in evaluating potential cost reductions that can be realized by the optimization of design and installation and construction sequences. The intent of this research is to see if this type of technology can be used in capacities similar to those currently filled by full-scale physical mockups and desktop simulations. Using a fully-immersive five sided virtual reality system, known as a CAVE, the authors constructed a series of virtual mockups that represented two next generation nuclear power plants, the Westinghouse AP-1000 and the Pebble Bed Modular Reactor (PBMR). These virtual mockups were then tested as a design tool to help locate and correct problem areas, to optimize the construction sequence, and to assist with familiarizing trades people with the performance of maintenance activities. A series of experiments were performed to assess the usefulness of these virtual mockups in accomplishing these tasks. (authors)

  6. Desalination of seawater: a nuclear solution

    International Nuclear Information System (INIS)

    Basta, H.

    2003-01-01

    1,4 milliard human beings do not benefit of sufficient clean water supply. The desalting of seawater is a valid and tested solution in terms of technology but the 2 physical processes involved, evaporation and reverse osmosis are energy-greedy. Only rich countries like Kuwait or Saudi-Arabia can afford producing massive volumes of fresh water from seawater. Today the total world capacity of desalting reaches 30 milliard m 3 a day with 10.000 operating units, half of which installed in middle-east countries. The use of nuclear energy is a solution to lower costs. In Aktau (Kazakhstan) a BN-350 fast reactor has been producing a 135 MW electrical output and 80.000 m 3 of fresh water a day for 27 years. In Japan about 10 desalting units have been coupled to nuclear power plants. A company (Eskom) based in South-Africa is developing a new concept of high temperature reactor: the PBMR (pebble bed modular reactor). The suitability of this reactor has been assessed for desalting and it appears that its main assets are its size: 165 MW electrical output (400 MW thermal output) and its Brayton cycle. Other characteristics such as the coolant (helium), the type of fuel (8% enriched uranium encapsulated in carbon), the low design and maintenance costs, the short building time (2 years) are important when considering issues like nuclear safety, non-proliferation and profitability. (A.C.)

  7. Contact detection acceleration in pebble flow simulation for pebble bed reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y.; Ji, W. [Department of Mechanical, Aerospace, and Nuclear Engineering Rensselaer, Polytechnic Institute, 110 8th street, Troy, NY 12180 (United States)

    2013-07-01

    Pebble flow simulation plays an important role in the steady state and transient analysis of thermal-hydraulics and neutronics for Pebble Bed Reactors (PBR). The Discrete Element Method (DEM) and the modified Molecular Dynamics (MD) method are widely used to simulate the pebble motion to obtain the distribution of pebble concentration, velocity, and maximum contact stress. Although DEM and MD present high accuracy in the pebble flow simulation, they are quite computationally expensive due to the large quantity of pebbles to be simulated in a typical PBR and the ubiquitous contacts and collisions between neighboring pebbles that need to be detected frequently in the simulation, which greatly restricted their applicability for large scale PBR designs such as PBMR400. Since the contact detection accounts for more than 60% of the overall CPU time in the pebble flow simulation, the acceleration of the contact detection can greatly enhance the overall efficiency. In the present work, based on the design features of PBRs, two contact detection algorithms, the basic cell search algorithm and the bounding box search algorithm are investigated and applied to pebble contact detection. The influence from the PBR system size, core geometry and the searching cell size on the contact detection efficiency is presented. Our results suggest that for present PBR applications, the bounding box algorithm is less sensitive to the aforementioned effects and has superior performance in pebble contact detection compared with basic cell search algorithm. (authors)

  8. Contact detection acceleration in pebble flow simulation for pebble bed reactor systems

    International Nuclear Information System (INIS)

    Li, Y.; Ji, W.

    2013-01-01

    Pebble flow simulation plays an important role in the steady state and transient analysis of thermal-hydraulics and neutronics for Pebble Bed Reactors (PBR). The Discrete Element Method (DEM) and the modified Molecular Dynamics (MD) method are widely used to simulate the pebble motion to obtain the distribution of pebble concentration, velocity, and maximum contact stress. Although DEM and MD present high accuracy in the pebble flow simulation, they are quite computationally expensive due to the large quantity of pebbles to be simulated in a typical PBR and the ubiquitous contacts and collisions between neighboring pebbles that need to be detected frequently in the simulation, which greatly restricted their applicability for large scale PBR designs such as PBMR400. Since the contact detection accounts for more than 60% of the overall CPU time in the pebble flow simulation, the acceleration of the contact detection can greatly enhance the overall efficiency. In the present work, based on the design features of PBRs, two contact detection algorithms, the basic cell search algorithm and the bounding box search algorithm are investigated and applied to pebble contact detection. The influence from the PBR system size, core geometry and the searching cell size on the contact detection efficiency is presented. Our results suggest that for present PBR applications, the bounding box algorithm is less sensitive to the aforementioned effects and has superior performance in pebble contact detection compared with basic cell search algorithm. (authors)

  9. The Nuclear Renaissance - Implications on Quantitative Nondestructive Evaluations

    International Nuclear Information System (INIS)

    Matzie, Regis A.

    2007-01-01

    The world demand for energy is growing rapidly, particularly in developing countries that are trying to raise the standard of living for billions of people, many of whom do not even have access to electricity. With this increased energy demand and the high and volatile price of fossil fuels, nuclear energy is experiencing resurgence. This so-called nuclear renaissance is broad based, reaching across Asia, the United States, Europe, as well as selected countries in Africa and South America. Some countries, such as Italy, that have actually turned away from nuclear energy are reconsidering the advisability of this design. This renaissance provides the opportunity to deploy more advanced reactor designs that are operating today, with improved safety, economy, and operations. In this keynote address, I will briefly present three such advanced reactor designs in whose development Westinghouse is participating. These designs include the advanced passive PWR, AP1000, which recently received design certification for the US Nuclear Regulatory Commission; the Pebble Bed Modular reactor (PBMR) which is being demonstrated in South Africa; and the International Reactor Innovative and Secure (IRIS), which was showcased in the US Department of Energy's recently announced Global Nuclear Energy Partnership (GNEP), program. The salient features of these designs that impact future requirements on quantitative nondestructive evaluations will be discussed. Such features as reactor vessel materials, operating temperature regimes, and new geometric configurations will be described, and mention will be made of the impact on quantitative nondestructive evaluation (NDE) approaches

  10. Pebble bed modular reactor fuel enrichment discrimination using delayed neutrons - HTR2008-58133

    International Nuclear Information System (INIS)

    Skoda, R.; Rataj, J.; Uhera, J.

    2008-01-01

    The Pebble Bed Modular Reactor (PBMR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor which utilise fuel in form of spheres that are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burn-up limit. When the reactor is started up for the first time, the lower-enriched start-up fuel is used, mixed with graphite spheres, to bring the core to criticality. As the core criticality is established and the start-up fuel is burned-in, the graphite spheres are progressively removed and replaced with more start-up fuel. Once it becomes necessary for maintaining power output, the higher enriched equilibrium fuel is introduced to the reactor and the start-up fuel is removed. During the initial run of the reactor it is important to discriminate between the irradiated startup fuel and the irradiated equilibrium fuel to ensure that only the equilibrium fuel is returned to the reactor. There is therefore a need for an on-line enrichment discrimination device that can discriminate between irradiated start-up fuel spheres and irradiated equilibrium fuel spheres. The device must also not be confused by the presence of any remaining graphite spheres. Due to it's on-line nature the device must accomplish the discrimination within tight time limits. Theoretical calculations and experiments show that Fuel Enrichment Discrimination based on delayed neutrons detection is possible. The paper presents calculations and experiments showing viability of the method. (authors)

  11. The Role of Exponential and PCTR Experiments at Hanford in the Design of Large Power Reactors; Roles Respectifs des Experiences Exponentielles et du Reacteur d'Etude des Constantes Physiques de Hanford dans les Etudes de Grands Reacteurs de Puissance; Znachenie ehksponentsial'nykh opytov i opytov na reaktore PCTR pri proektirovanii bol'shikh ehnergeticheskikh reaktorov v khehnforde; Papel de los Experimentos Exponenciales y del Reactor PCTR de Hanford en el Proyecto de Grandes Reactores de Potencia

    Energy Technology Data Exchange (ETDEWEB)

    Heineman, R. E. [General Electric Company, Richland, WA (United States)

    1964-02-15

    introduccion de cambios en el manejo de los reactores ya existentes, como instrumentos de investigacion en la esfera de la fisica de los reactores y como medio de ensenanza. Compara tambien los capitales invertidos en esas instalaciones y los gastos de explotacion. Describe el perfeccionamiento de nuevas tecnicas experimentales que estas instalaciones permiten aplicar con miras a satisfacer la demanda de nuevos datos experimentales. Es menester tener presentes todos estos datos para poder predecir la evolucion de las necesidades y las tendencias futuras en el empleo de estas instalaciones para los estudios de los reactores de potencia. La memoria describe sucintamente el reactor para el estudio de constantes fisicas e indica la manera en que se piensa utilizarlo en el marco de esa evolucion. (author) [Russian] V Hjen- fordskih laboratorijah v techenie pochti 15 let provodjatsja jeksponencial'nye reaktornye iz- merenija na grafito-uranovyh reshetkah. Hotja rezul'taty jetih opytov ispol'zovalis' dlja opredelenija laplasianov predlagaemyh proizvodjashhih reaktorov, oni takzhe sodejstvovali razvitiju ponimanija fiziki reaktorov jetih sistem. Davno priznano, chto poleznost' kri- ticheskogo opyta ogranichena vvidu ego bol'shogo masshtaba i nedostatochnoj chuvstvitel'nosti v otnoshenii nebol'shih lokalizovannyh narushenij sistemy. Zatem mysl' byla napravlena na sozdanie cel'nogo opytnogo reaktora, v kotorom bylo by svedeno do minimuma kolichest- vo materialov, neobhodimyh dlja poluchenija nuzhnyh dannyh. Jeta popytka privela k postrojke usovershenstvovannoj kriticheskoj ustanovki s neskol'kimi zonami reaktora dlja izmerenija fizicheskih konstant PCTR. Ustanovka ispol'zuetsja dlja okazanija sodejstvija pri razrabot- ke proekta po fizike reaktorov dlja neskol'kih jenergeticheskih reaktorov. Krome togo,re- aktor RSTNjavljaetsja ustanovkoj obshhego naznachenija dlja provedenija izmerenij poperechnyh sechenij na reaktore i dlja opredelenija differencial'nyh i integral'nyh fizicheskih para

  12. RB Research nuclear reactor, Annual report for 1994, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1994. godini, I - III

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Milosevic, M; Pesic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia); Marinkovic, P [Elektrotehnicki fakultet, beograd (Yugoslavia); Kocic, A; Ilic, R; Dasic, N; Ljubenov, V; Petronijevic, M; Jevremovic, M [Institute of Nuclear Sciences Vinca, Belgrade (Serbia)

    1994-12-15

    Report on RB reactor operation during 1994 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization as well as operation of the VAX-8250 computer.

  13. RB research nuclear reactor, Annual report for 1989, I - III; Istrazivacki nukleani reaktor RB (Izvestaj o radu u 1989. godini), I - III

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Pesic, M; Hadimahmutovic, N; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1989-12-15

    This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989.

  14. RB Research nuclear reactor, Annual report for 1995, I-IV; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1995. godini, I-IV

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Milosevic, M; Pesic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia); Marinkovic, P [Elektrotehnicki fakultet, Beograd (Yugoslavia); Ilic, R; Dasic, N; Milovanovic, S; Ljubenov, V; Petronijevic, M; Jevremovic, M [Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1995-12-15

    Report on RB reactor operation during 1995 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor.

  15. RB research nuclear reactor, Annual report for 1984, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1984. godini, I - III

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1984-07-01

    The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source.

  16. RB research nuclear reactor - Annual report for 1986, I - III; Istrazivacki nuklearni reaktor RB (Izvestaj o radu u 1986. godini), I-III

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Jevremovic, M; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1987-07-01

    This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986. Izvestaj pokazuje podatke o radu reaktora RB u toku 1986. godine, stanje reaktorske opreme, podatke o angazovanom osoblju na reaktoru i datoteku sa podacima o eksperimentalnoj i drugoj opremi reaktora RB. Sastoji se od 3 dela: tehnicki opis, pogon i rad reaktora, oprema postrojenja i njeno odrzavanje, koriscenje reaktora u 1986. godini.

  17. Measurements with a Pulsed and Modulated Source in a Reactor; Mesures au Moyen d'une Source Pulsee et Modulee dans un Reacteur; Izmereniya v reaktore s pomoshch'yu impul'snogo i moduliruemogo is tochnika; Mediciones Efectuadas en Reactor con una Fuente Pulsada y Modulada

    Energy Technology Data Exchange (ETDEWEB)

    Rotter, W. [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

    1965-10-15

    analizador multicanal. Como el flujo del generador es perfectamente sinusoidal, la respuesta del reactor puede integrarse en cada cuarto de perfodo, puesto que el circuito de medicion esta gobernado por el generador; por consiguiente, el tiempo de medicion es minimo. Los datos registrados sobre cinta perforada se analizan con ayuda de una calculadora numerica. (author) [Russian] Generator, vyhod nejtronov kotorogo izmenjaetsja v zavisimosti ot funkcii kakogo-to vremeni, byl razrabotan v issledovatel'skih laboratorijah Filipsa. Ego prakticheskaja pol'za v oblasti fiziki reaktorov byla prodemonstrirovana na serii izmerenij, provedennyh v reaktore BR-O 2 v podkriticheskom sostojanii. Horoshaja ustojchivost', vozmozhnost' proizvodit' rezkie izmenenija intensivnosti nejtronnogo potoka, impul'sirovat' vyhod ili modulirovat' ego sinusoidal'no, - vse jeto delaet takoj generator ochen' gibkim. On pozvoljaet ustanavlivat' reaktivnost' ({rho} = {Delta}k/{beta}) i vremja zhizni nejtronov ( Script-Small-L /{beta}) po razlichnym nezavisimym metodam. Tochnoe sravnenie jetih metodov vozmozhno, poskol'ku poslednie mogut byt' ispol'zovany bez izmenenija uslovij izmerenija. Ustanovleno: 1) {rho} na osnove zapazdyvajushhih nejtronov putem mgnovennogo umen'shenija vyhoda nejtronov; 2) {rho} na osnove mgnovennyh nejtronov putem ispol'zovanija impul'sov nejtronov; 3) Script-Small-L /{beta} putem soedinenija 1) i 2) dlja 0,5 $ < {rho} < 2 $ ; 4) Script-Small-L /{beta} na osnove peredatochnoj funkcii reaktora dlja moduliruemogo istochnika. Obsuzhdajutsja peredatochnye funkcii dlja oscilljatora reaktivnosti i dlja sinusoidal'no moduliruemogo istochnika. Pokazano, chto izmerenie Script-Small-L /{beta} vozmozhno dlja 0,1 $< {rho} < $ s uchetom primenenija moduliruemogo istochnika. Tot zhe metod takzhe daet reaktivnost' s pomoshh'ju otnoshenija mgnovennyh nejtronov k zapazdyvajushhim nejtronam dlja optimal'noj chastoty, prichem na praktike jeto proishodit nezavisimo ot dannyh, otnosjashhihsja k

  18. Thermal Shock Tests on UO{sub 2} Small Spheres; Essais de choc thermique sur des elements spheriques de UO{sub 2}; Ispytaniya nebol'shikh sharikov iz UO{sub 2} teplovykh udarom; Ensayo de pequenas esferas de UO{sub 2} por choque.termico

    Energy Technology Data Exchange (ETDEWEB)

    Perona, G.; Brutto, E.; Galbusera, U.; Palladino, G.; Sesini, R. [Centro Informazioni Studi Esperienze, Milan (Italy)

    1963-11-15

    If UO{sub 2} small spheres are used as fuel in a reactor in contact with the cooler, it is necessary to know the maximum value of the thermal stress, due to the work conditions in the reactor, which the small spheres are able to withstand without breaking. These conditions can be calculated if the physical properties of the material are known. Owing to the considerable number of properties involved, and in consideration of the uncertainty which always exists in each of them, it is preferable to test directly the spheres, submitting them to the same kind of stresses that they undergo in thereactor. In this work a thermal shock method for the small spheres has been studied, while conditions are indicated in which this method can reproduce stress conditions directly comparable with those existing in the reactor. As for small spheres, the difficulty consists in producing coolings with very high values of the coefficient of surface heat transfer. The experimental methods are described and the results obtained are indicated. The application of this method seems to be very interesting particularly in the field of the technological research for improving the characteristics of the UO{sub 2} small spheres by means of additives. In fact it allows the control of the total interesting effect with a single measurement. (author) [French] Si l'on veut utiliser comme combustible dans un reacteur des elements spheriques de UO{sub 2} en contact avec le refroidisseur, il faut au prealable determiner la valeur maximum de la contrainte thermique - due aux conditions regnant dans le reacteura laquelle les elements sont capables de resister sans se fissurer. Il est possible de calculer ces conditions si l'on connait les proprietes physiques du materiau utilise. En raison du nombre important des proprietes a prendre en consideration, et compte tenu de l'incertitude qui existe toujours pour chacune d'elles, il est preferable de faire des essais thermiques en soumettant directement les

  19. Research Project 'RB research nuclear reactor' (operation and maintenance), Final report; Naucnoistrazivacki projekt 'Istrazivacki nuclearni reaktor RB, (pogon i odrzavanje), Zavrsni elaborat projekta

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    This final report covers operation and maintenance activities at the RB reactor during period from 1981-1985. First part covers the RB reactor operation, detailed description of reactor components, fuel, heavy water, reactor vessel, cooling system, equipment and instrumentation, auxiliary systems. It contains data concerned with dosimetry and radiation protection, reactor staff, and financial data. Second part deals maintenance, regular control and testing of reactor equipment and instrumentation. Third part is devoted to basic experimental options and utilization of the RB reactor including training.

  20. Control of a reactor for conditioning of biogenic waste materials. Final report; Steuerung eines Reaktors zur Aufbereitung von Abfaellen mit biogenen Bestandteilen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Bartha, B.; Brummack, J.; Kloeden, W.

    2002-10-01

    Mechanical-biological waste treatment implies biological drying and mechanical separation; it produces a pollutant-free fuel fraction with good calorific value and inorganic fractions available for recycling. Apart from some prototypes, mostly static aerobic reactors are used. For moist residues, dynamic reactors may be interesting as they permit coupling of thermal and mechanical processes, i.e. parallel biological drying and mechanical separation. A discontinuous rotary kiln reactor for biological drying of residues was developed in this project. (orig.) [German] Die mechanisch-biologische Aufbereitung von Restabfaellen (MBA) hat sich zu einem Systembaustein innerhalb der Restabfallentsorgung entwickelt. Das Hauptanwendungsziel besteht darin, durch die Kombination von biologischer Trocknung mit mechanischen Stofftrennverfahren eine heizwertreiche und schadstoffentfrachtete Fraktion mit Brennstoffeigenschaften, sowie verwertbare anorganische Fraktionen aus dem Restabfall zu gewinnen. Die Ausgangssituation zeigte, dass, abgesehen von einigen prototypischen Anlagen, in der MBA-Technologie hauptsaechlich statische, aerob arbeitende Reaktoren (Rottetunnel und Rotteboxen) fuer den biologischen Schritt eingesetzt werden. Fuer feuchte Restabfaelle erscheint es interessant, dynamische Reaktoren einzusetzen, die die Kopplung thermischer und mechanischer Prozesse erlauben, die also die biologische Trocknung und den mechanischen Aufschluss parallel zulassen. Im Rahmen des Projektes wurde die Steuerung eines diskontinuierlich betriebenen Drehrohrreaktors zur biologischen Trocknung von Restabfaellen entwickelt. (orig.)

  1. The Siemens-Argonaut reactor as a driver zone for a high-temperature reactor cell. Der Siemens-Argonaut-Reaktor als Treiberzone fuer eine Hochtemperaturreaktorzelle

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H; Schuerrer, F; Ninaus, W; Oswald, K; Rabitsch, H; Kreiner, H [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik; Neef, R D [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Reaktorentwicklung

    1984-12-15

    To enable a validation of neutron physics calculation methods for pebble bed reactors the inner reflector of an Argonaut research reactor was substituted by a full of about 1200 fuel elements of the AVR-Juelich type. The report describes the measuring instruments and the reactor physical layout of the arrangement by the code packages GAMTEREX, ZUT-D.G.L. and MUPO. Comparison of calculated reaction rates with measurements show good agreement. Application of the codes to high-temperature reactors in abnormal states is envisaged. (Author, translated by G.Q.)

  2. Preliminary design of RDE feedwater pump impeller

    International Nuclear Information System (INIS)

    Sri Sudadiyo

    2018-01-01

    Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Experimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MW th , and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm. (author)

  3. Irradiation effects on Zr-2.5Nb in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

  4. Operation and maintenance of the RA nuclear reactor for 1977, Report Annex I; Rad i iskoriscenost reaktora RA u 1977. godini, izvestaj Prilog I

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Stanic, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1977-12-15

    RA reactor operation plan was fulfilled, meaning 28583 MWh. In addition to 183 days of operation at full power, during 1977 the reactor was operated for 14 days at lower power level according to the demand of users. The utilization level of rector irradiation capability (neutron flux and time of operation) was 14%. This annex includes detailed statistical data about reactor operation, utilization, power level, savings concerned with new 80% enriched fuel. All the 9 vertical experimental channels were used for irradiation in the reactor core. Digression from the action plan were caused by refueling and demand od the users. There have been 11 safety shutdowns, of which 6 caused by power cuts, 4 due to failure of the instruments, and 1 due to earthquake in March 1977. [Serbo-Croat] Planirani rad reaktora na nominalnoj snazi ostvaren je u iznosu od 28583 MWh. U toku 1977. godine reaktor je radio 14 dana na manjim snagama i u posebnom eksperimentalnom rezimu na zahtev korisnika. Iskoriscenost kapaciteta reaktora za ozracivanje uzoraka (na bazi neutronskog fluksa i vremena rada reaktora) bila je 41%. Ovaj izvestaj sadrzi detaljne statisticke podatke o radu i iskoriscenosti reaktora, podatke o ustedi goriva prelaskom na 80% obogaceno gorivo. Korisceno je svih 9 vertikalnih eksperimentalnih kanala u aktivnoj zoni. Uzroci odstupanja od plana rada osim zamene goriva bili su zahtevi korisnika. Bilo je 11 sigurnosnih zaustavljanja, 6 usled nestanka napona, 4 usled kvarova opreme i instrumentacije, i 1 put usled zemljotresa u martu.

  5. Upgrade of Control and Protection System of the Ignalina Nuclear Power Plant Units 1 and 2

    International Nuclear Information System (INIS)

    Wright, Ronald E.; Fletcher, Norman; Sidnev, Victor E.; Bickel, John H.; Vianello, Aldo; Pearsall, Raymond D.

    2003-01-01

    The Ignalina nuclear power plant (NPP) Units 1 and 2 are Soviet-designed, RBMK (Reaktor Bolshoi Moschnosti Kipyashchiy), channelized, large power-type reactors. The original-design electrical capacity for each unit was 1500 MW. Unit 1 began operating in 1983, and Unit 2 was started up in 1987. In 1994, the government of Lithuania agreed to accept grant support for the Ignalina NPP Safety Improvement Program with funding supplied by the Nuclear Safety Account of the European Bank for Reconstruction and Development (EBRD). As conditions for receiving this funding, the Ignalina NPP agreed to prepare a comprehensive safety analysis report that would undergo independent peer review after it was issued. The EBRD Safety Panel oversaw preparation and review of the report. In 1996, the safety analysis report for Unit 1 was completed and delivered to the EBRD. Part of the analyses covered anticipated transients without scram (ATWS). The analysis showed that some ATWS scenarios could lead to unacceptable consequences in <1 min. The EBRD Safety Panel recommended to the government of Lithuania that the Ignalina NPP develop and implement a program of compensatory measures for the control and protection system before the unit would be allowed to return to operation following its 1998 maintenance outage. A compensatory control and protection system that would mitigate the unacceptable consequences was designed, procured, manufactured, tested, and installed. The project was funded by U.S. Department of Energy

  6. Degradasi Zat Warna Pada Limbah Cair Industri Tekstil Dengan Metode Fotokatalitik Menggunakan Nanokomposit Tio2 – Zeolit

    Directory of Open Access Journals (Sweden)

    Siti Naimah

    2014-10-01

    Full Text Available Telah dilakukan penelitian degradasi zat warna pada limbah cair industri tekstil menggunakan metode fotokatalitik dengan penambahan nanokomposit TiO2 - zeolit. Tujuan penelitian ini untuk mengetahui efektifitas kemampuan nanokomposit dalam mendegradasi zat warna serta parameter-parameter yang ditetapkan dalam Peraturan Pemerintah Nomor 82 Tahun 2001 tentang pengelolaan kualitas air dan pengendalian pencemaran air. Zeolit alam diaktivasi terlebih dahulu sebelum dikompositkan dengan TiO2. Perbandingan TiO2 : zeolit yang digunakan pada pembuatan nanokomposit adalah 100:0, 20:80, 40:60, 50:50, 60:40, dan 0:100. Percobaan pendahuluan dilakukan dengan menggunakan limbah cair tekstil buatan yang dibuat dari pewarna Synolon yellow S- G6LS (untuk warna kuning dan B/Blue R 150% special (untuk warna biru, sedangkan limbah cair industri tekstil diambil dari salah satu industri di Bogor. Waktu degradasi zat warna dilakukan dalam reaktor fotokatalitik selama 180 menit. Pada perbandingan TiO2 : zeolit 40:60 didapatkan degradasi zat warna tekstil buatan berwarna kuning maksimal adalah 99,9 % dan zat warna tekstil buatan berwarna biru maksimal 99,8%. Analisis warna menggunakan spektrofotometer dan HPLC. Nanokomposit TiO2 : zeolit 40 : 60 merupakan perbandingan optimal sehingga digunakan pada uji coba limbah cair industri tekstil. Degradasi maksimal warna kuning dengan pengolahan fotokatalitik yang ditambahkan nanokomposit pada limbah cair industri tekstil sebesar 98,4%, sedangkan untuk parameter uji zat organik, TSS, TDS, BOD, COD, dan lemak/minyak diperoleh nilai di bawah baku mutu yang dipersyaratkan. 

  7. The evolution of the break preclusion concept for nuclear power plants in Germany

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, H. [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany)

    1997-04-01

    In the updating of the Guidelines for PWR`s of the {open_quotes}Reaktor-Sicherheitskommission{close_quotes} (RSK) in 1981 the requirements on the design have been changed with respect to the postulated leaks and breaks in the primary pressure boundary. The major change was a revision in the requirements for pipe whip protection. As a logical consequence of the {open_quotes}concept of basic safety{close_quotes} a guillotine type break or any other break type resulting in a large opening is not postulated any longer for the calculation of reaction and jet forces. As an upper limit for a leak an area of 0, 1 A (A = open cross section of the pipe) is postulated. This decision was based on a general assessment of the present PWR system design in Germany. Since then a number of piping systems have been requalified in the older nuclear power plants to comply with the break preclusion concept. Also a number of extensions of the concept have been developed to cover also leak-assumptions for branch pipes. Furthermore due considerations have been given to other aspects which could contribute to a leak development in the primary circuit, like vessel penetrations, manhole covers, flanges, etc. Now the break preclusion concept originally applied to the main piping has been developed into an integrated concept for the whole pressure boundary within the containment and will be applied also in the periodic safety review of present nuclear power plants.

  8. A systems CFD model of a packed bed high temperature gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Du Toit, C.G.; Rousseau, P.G.; Greyvenstein, G.P.; Landman, W.A.

    2006-01-01

    The theoretical basis and conceptual formulation of a comprehensive reactor model to simulate the thermal-fluid phenomena of the PBMR reactor core and core structures is given. Through a rigorous analysis the fundamental equations are recast in a form that is suitable for incorporation in a systems CFD code. The formulation of the equations results in a collection of one-dimensional elements (models) that can be used to construct a comprehensive multi-dimensional network model of the reactor. The elements account for the pressure drop through the reactor; the convective heat transport by the gas; the convection heat transfer between the gas and the solids; the radiative, contact and convection heat transfer between the pebbles and the heat conduction in the pebbles. Results from the numerical model are compared with that of experiments conducted on the SANA facility covering a range of temperatures as well as two different fluids and different heating configurations. The good comparison obtained between the simulated and measured results show that the systems CFD approach sufficiently accounts for all of the important phenomena encountered in the quasi-steady natural convection driven flows that will prevail after critical events in a reactor. The fact that the computer simulation time for all of the simulations was less than three seconds on a standard notebook computer also indicates that the new model indeed achieves a fine balance between accuracy and simplicity. The new model can therefore be used with confidence and still allow quick integrated plant simulations. (authors)

  9. South Africa's opportunity to maximise the role of nuclear power in a global hydrogen economy

    Energy Technology Data Exchange (ETDEWEB)

    Greyvenstein, R. [Pebble Bed Modular Reactor (PBMR) (Pty) Ltd. (South Africa)], E-mail: renee.greyvenstein@pbmr.co.za; Correia, M. [Pebble Bed Modular Reactor (PBMR) (Pty) Ltd. (South Africa)], E-mail: michael.correia@pbmr.co.za; Kriel, W. [Pebble Bed Modular Reactor (PBMR) (Pty) Ltd. (South Africa)], E-mail: willem.kriel@pbmr.us

    2008-11-15

    Global concern for increased energy demand, increased cost of natural gas and petroleum, energy security and environmental degradation are leading to heightened interest in using nuclear energy and hydrogen to leverage existing hydrocarbon reserves. The wasteful use of hydrocarbons can be minimised by using nuclear as a source of energy and water as a source of hydrogen. Virtually all hydrogen today is produced from fossil fuels, which give rise to CO{sub 2} emissions. Hydrogen can be cleanly produced from water (without CO{sub 2} pollution) by using nuclear energy to generate the required electricity and/or process heat to split the water molecule. Once the clean hydrogen has been produced, it can be used as feedstock to fuel cell technologies, or in the nearer term as feedstock to a coal-to-liquids process to produce cleaner synthetic liquid fuels. Clean liquid fuels from coal - using hydrogen generated from nuclear energy - is an intermediate step for using hydrogen to reduce pollution in the transport sector; simultaneously addressing energy security concerns. Several promising water-splitting technologies have been identified. Thermo-chemical water-splitting and high-temperature steam electrolysis technologies require process temperatures in the range of 850 deg. C and higher for the efficient production of hydrogen. The pebble bed modular reactor (PBMR), under development in South Africa, is ideally suited to generate both high-temperature process heat and electricity for the production of hydrogen. This paper will discuss South Africa's opportunity to maximise the use of its nuclear technology and national resources in a global hydrogen economy.

  10. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR and GEN IV

    International Nuclear Information System (INIS)

    O'Donnell, William J.; Griffin, Donald S.

    2007-01-01

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  11. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

    Energy Technology Data Exchange (ETDEWEB)

    William J. O’Donnell; Donald S. Griffin

    2007-05-07

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  12. Report on results of the Experts from the Boris Kidric Institute to the Institute for theoretical and experimental physics in Moscow - Operational Report; Radni izvestaj - Izvestaj o rezultatima posete strucnjaka iz IBK Institutu za teorijsku i eksperimentalnu fiziku u Moskvi

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Petrovic, M; Cupac, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1976-12-15

    .) upoznavanje eksploatacionih karakteristika reaktora TVRS sa 80% obogacenim gorivom, kao i iskustva u vezi sa koriscenjem tog goriva u reaktoru TVRS u Moskvi; b.) upoznavanje sa elementima sigurnosnih analiza u vezi sa uvodjenjem i kriscenjem tog goriva u reaktoru TVRS kao i c.) razmena gledista o programu prevodjenja reaktora RA na novo 80% obogaceno gorivo. U izvestaju su prikazani rezultati ove posete. Osnovni zakljucci do kojih se doslo jesu: a.) poseta je bila korisna i pored toga sto na neka pitanja nije dobijen odgovor; b.) posle obrade dobijenih informacija, posle sticanja prvih iskustava o koriscenju novog goriva u reaktoru RA kao i posle donosenja definitivnog suda o ravnoteznom rezimu reaktora RA sa novim gorivom, bilo bi veoma korisno da se izvrsi jos jedna poseta Institutu u Moskvi; c.) novo disperziono gorivo je mnogo pouzdanije u eksploataciji od starog 2% obogacenog metalnog goriva (kakvo je sada u reaktoru RA); d.) nije bilo primedbi sa aspekta sigurnosti na nas program prevodjenja (prelazni rezim) reaktora RA na novo gorivo; e.) zatrazice se dopunske informacije o maksimalnom dopustivom stepenu izgaranja 80% obogacenog disperzionog goriva i po potrebi izvrsice se blagovremena revizija predvidjenog optimalnog ravnoteznog rezima izgaranja za reaktor RA (baziranog na maksimalnom izgaranju goriva koje omogucuju parametri reaktora RA); f.) metodologija termickog proracuna gorivnih kanala, koja se primenjuje na reaktor RA ocenjena je kao ekvivalentna postupku koji se primenjuje za proracun kanala reaktora TVRS u Moskvi (author)

  13. DESAIN AWAL TURBIN UAP TIPE AKSIAL UNTUK KONSEP RGTT30 BERPENDINGIN HELIUM

    Directory of Open Access Journals (Sweden)

    Sri Sudadiyo

    2016-06-01

    Full Text Available ABSTRAK DESAIN AWAL TURBIN UAP TIPE AKSIAL UNTUK KONSEP RGTT30 BERPENDINGIN HELIUM. Konsep reaktor daya nuklir yang dikembangkan merupakan jenis reaktor berpendingin gas dengan temperatur tinggi (RGTT. Gas yang digunakan untuk mendinginkan teras RGTT adalah helium. Konsep RGTT ini dapat menghasilkan daya termal 30 MWth sehingga dinamakan RGTT30. Temperatur helium mampu mencapai 700 °C ketika keluar dari teras RGTT30 dan digunakan untuk memanaskan air di dalam steam generator hingga mencapai temperatur 435 °C. Steam generator dihubungkan dengan turbin uap yang dikopel dengan generator listrik untuk membangkitkan daya 7,27 MWe. Uap yang keluar dari turbin dilewatkan kondensor untuk mencairkan uap menjadi air. Rangkaian komponen dari steam generator, turbin, dan kondensor dinamakan sistem turbin uap. Turbin terdiri dari sudu-sudu yang dimaksudkan untuk mengubah tenaga uap kedalam tenaga mekanis berupa putaran. Efisiensi turbin merupakan parameter yang harus diperhatikan dalam sistem turbin uap ini. Tujuan dari makalah ini adalah untuk mengusulkan sudu tipe aksial dan untuk menganalisa perbaikan efisiensi turbin. Metode yang digunakan yaitu aplikasi prinsip termodinamika yang berhubungan dengan konservasi energi dan massa. Perangkat lunak Cycle-Tempo dipakai untuk mendapatkan parameter termodinamika dan untuk mensimulasikan sistem turbin uap berbasis RGTT30. Pertama, dibuat skenario dalam simulasi sistem turbin uap untuk mengetahui efisiensi dan laju aliran massa uap yang diperoleh nilai optimal 87,52 % dan 8,759 kg/s pada putaran 3000 rpm. Kemudian, turbin uap diberi sudu tipe aksial dengan diameter tip 1580 mm dan panjang 150 mm. Hasil yang diperoleh adalah nilai efisiensi turbin uap naik menjadi 88,3 % pada putaran konstan (3000 rpm. Penambahan nilai efisiensi turbin sebesar 0,78 % menunjukkan peningkatan kinerja RGTT30 secara keseluruhan. Kata kunci: Tipe aksial, turbin uap, RGTT30   ABSTRACT PRELIMINARY DESIGN ON STEAM TURBINE OF AXIAL TYPE

  14. A CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR FOR IN VITRO AND IN VIVO TEST OF BORON NEUTRON CAPTURE THERAPY

    Directory of Open Access Journals (Sweden)

    Nina Fauziah

    2015-03-01

    Full Text Available Studies were carried out to design a collimator which results in epithermal neutron beam for IN VITRO and IN VIVO of Boron Neutron Capture Therapy (BNCT at the Kartini research reactor by means of Monte Carlo N-Particle (MCNP codes. Reactor within 100 kW of thermal power was used as the neutron source. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 2 cm thick of Bi as γ-ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 1 to 5 cm varied aperture size, epithermal neutron beam with maximum flux of 7.65 x 108 n.cm-2.s-1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 1.76 x 10-13 Gy.cm2.n-1 and 1.32 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.008, and maximum directionality of 0.73. It did not fully pass the IAEA’s criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108 n.cm-2.s-1. When it was assumed that the graphite inside the thermal column was not discharged but only the part which was going to be replaced by the collimator, the performance of the collimator became better within the positive effect from the surrounding graphite that the beam resulted passed all criteria with epithermal neutron flux up to 1.68 x 109 n.cm-2.s-1. Keywords: design, collimator, epithermal neutron beam, BNCT, MCNP, criteria   Telah dilakukan penelitian tentang desain kolimator yang menghasilkan radiasi netron epitermal untuk uji in vitro dan in vivo pada Boron Neutron Capture Therapy (BNCT di Reaktor Riset Kartini dengan menggunakan program Monte

  15. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-04-15

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  16. Evaluation of High Temperature Gas Cooled Reactor Performance: Benchmark Analysis Related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA Critical Facility

    International Nuclear Information System (INIS)

    2013-04-01

    The IAEA has facilitated an extensive programme that addresses the technical development of advanced gas cooled reactor technology. Included in this programme is the coordinated research project (CRP) on Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance, which is the focus of this TECDOC. This CRP was established to foster the sharing of research and associated technical information among participating Member States in the ongoing development of the HTGR as a future source of nuclear energy. Within it, computer codes and models were verified through actual test results from operating reactor facilities. The work carried out in the CRP involved both computational and experimental analysis at various facilities in IAEA Member States with a view to verifying computer codes and methods in particular, and to evaluating the performance of HTGRs in general. The IAEA is grateful to China, the Russian Federation and South Africa for providing their facilities and benchmark programmes in support of this CRP.

  17. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    International Nuclear Information System (INIS)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook

    2007-08-01

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the modeling

  18. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  19. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavron, Barak, E-mail: btavron@bgu.ac.il [Planning, Development and Technology Division, Israel Electric Corporation Ltd., P.O. Box 10, Haifa 31000 (Israel); Shwageraus, Eugene, E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)

    2016-10-15

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  20. Pebble bed reactor fuel cycle optimization using particle swarm algorithm

    International Nuclear Information System (INIS)

    Tavron, Barak; Shwageraus, Eugene

    2016-01-01

    Highlights: • Particle swarm method has been developed for fuel cycle optimization of PBR reactor. • Results show uranium utilization low sensitivity to fuel and core design parameters. • Multi-zone fuel loading pattern leads to a small improvement in uranium utilization. • Thorium mixes with highly enriched uranium yields the best uranium utilization. - Abstract: Pebble bed reactors (PBR) features, such as robust thermo-mechanical fuel design and on-line continuous fueling, facilitate wide range of fuel cycle alternatives. A range off fuel pebble types, containing different amounts of fertile or fissile fuel material, may be loaded into the reactor core. Several fuel loading zones may be used since radial mixing of the pebbles was shown to be limited. This radial separation suggests the possibility to implement the “seed-blanket” concept for the utilization of fertile fuels such as thorium, and for enhancing reactor fuel utilization. In this study, the particle-swarm meta-heuristic evolutionary optimization method (PSO) has been used to find optimal fuel cycle design which yields the highest natural uranium utilization. The PSO method is known for solving efficiently complex problems with non-linear objective function, continuous or discrete parameters and complex constrains. The VSOP system of codes has been used for PBR fuel utilization calculations and MATLAB script has been used to implement the PSO algorithm. Optimization of PBR natural uranium utilization (NUU) has been carried out for 3000 MWth High Temperature Reactor design (HTR) operating on the Once Trough Then Out (OTTO) fuel management scheme, and for 400 MWth Pebble Bed Modular Reactor (PBMR) operating on the multi-pass (MEDUL) fuel management scheme. Results showed only a modest improvement in the NUU (<5%) over reference designs. Investigation of thorium fuel cases showed that the use of HEU in combination with thorium results in the most favorable reactor performance in terms of

  1. Evaluation, Comparison and Optimization of the Compact Recuperator for the High Temperature Gas-Cooled Reactor (HTGR) Helium Turbine System

    International Nuclear Information System (INIS)

    Hao Haoran; Yang Xiaoyong; Wang Jie; Ye Ping; Yu Xiaoli; Zhao Gang

    2014-01-01

    Helium turbine system is a promising method to covert the nuclear power generated by the High Temperature Gas Cooled Reactor (HTGR) into electricity with inherent safety, compact configuration and relative high efficiency. And the recuperator is one of the key components for the HTGR helium turbine system. It is used to recover the exhaust heat out of turbine and pass it to the helium from high pressure compressor, and hence increase the cycle’s efficiency dramatically. On the other hand, the pressure drop within the recuperator will reduce the cycle efficiency, especially on low pressure side of recuperator. It is necessary to optimize the design of recuperator to achieve better performance of HTGR helium turbine system. However, this optimization has to be performed with the restriction of the size of the pressure vessel which contains the power conversion unit. This paper firstly presents an analysis to investigate the effects of flow channel geometry, recuperator’s power and size on heat transfer and pressure drop. Then the relationship between the recuperator design and system performance is established with an analytical model, followed by the evaluations of the current recuperator designs of GT-MHR, GTHTR300 and PBMR, in which several effective technical measures to optimize the recuperator are compared. Finally it is found that the most important factors for optimizing recuperator design, i.e. the cross section dimensions and tortuosity of flow channel, which can also be extended to compact intermediate heat exchangers. It turns out that a proper optimization can increase the cycle’s efficiency by 1~2 percentage, which could also raise the economy and competitiveness of future commercial HTGR plants. (author)

  2. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

    RA research reactor is thermal heavy water moderated and cooled reactor. Metal uranium 2% enriched fuel elements were used at the beginning of its operation. Since 1976, 80% enriched uranium oxide dispersed in aluminium fuel elements were gradually introduced into the core and are the only ones presently used. Reactor core is cylindrical, having diameter 40 cm and 123 cm high. Reaktor core is made up of 82 fuel elements in aluminium channels, lattice is square, lattice pitch 13 cm. Reactor vessel is cylindrical made of 8 mm thick aluminium, inside diameter 140 cm and 5.5 m high surrounded with neutron reflector and biological shield. There is no containment, the reactor building is playing the shielding role. Three pumps enable circulation of heavy water in the primary cooling circuit. Degradation of heavy water is prevented by helium cover gas. Control rods with cadmium regulate the reactor operation. There are eleven absorption rods, seven are used for long term reactivity compensation, two for automatic power regulation and two for safety shutdown. Total anti reactivity of the rods amounts to 24%. RA reactor is equipped with a number of experimental channels, 45 vertical (9 in the core), 34 in the graphite reflector and two in the water biological shield; and six horizontal channels regularly distributed in the core. This volume include detailed description of systems and components of the RA reactor, reactor core parameters, thermal hydraulics of the core, fuel elements, fuel elements handling equipment, fuel management, and experimental devices [sr

  3. {sup 137}CS-determination in game meat from some hunting areas in lower Austria; {sup 137}Cs-Bestimmungen im Wildfleisch aus einigen niederoesterreichischen Jagdrevieren

    Energy Technology Data Exchange (ETDEWEB)

    Ayromlou, S. [Inst. fuer Anorganische Chemie der Univ. Wien (Austria); Tataruch, F. [Forschungsinstitut fuer Wildtierkunde und Oekologie der Veterinaermedizinischen Univ., Wien (Austria)

    2001-07-01

    In 1986, the contamination of some regions in Austria by {sup 137}Cs, due to the Chernobyl accident was relatively high. Among other {sup 137}Cs is taken up by people by the consumption of game. In an area of Lower Austria which is relatively heavily contaminated by Chernobyl fallout since 1986 the {sup 137}Cs-contamination of meat of game-animals was measured every year. Clear differences arose in the load of the single game species whose causes just like the temporal changes are discussed. The highest {sup 137}Cs activity concentration was with 5243 Bq/kg measured on a wild boar. With this activity concentration an annual effective dose of only 0,06 mSv can be estimated for an annual average consumption of one kilogram meat of wild boar. (orig.) [German] Druch den Reaktor-Unfall in Tschernobyl wurde Oesterreich gebietsweise relativ stark mit {sup 137}Cs kontaminiert. Unter anderem gelangt {sup 137}Cs durch den Verzehr von Wildfleisch in den Koerper der Menschen. Daher wurden in einem niederoesterreichischen Gebiet, das 1986 durch Fallout verhaeltnismaessig stark kontaminiert worden war seither jaehrlich erlegte Wildtiere auf den {sup 137}Cs-Gehalt ihres Fleisches hin untersucht. Dabei ergaben sich deutliche Unterschiede in der Belastung der einzelnen Wildarten, deren Ursachen ebenso wie die zeitlichen Veraenderungen diskutiert werden. Die hoechste {sup 137}Cs-Aktivitaetskonzentration wurde mit 5243 Bq/kg bei einem Wildschwein gemessen. Mit einen durchschnittlichen Verzehr von 1 kg Wildschweinfleisch pro Jahr kann daraus eine maximale Effektivdosis von nur 0,06 mSv/Jahr abgeschaetzt werden. (orig.)

  4. Heavy water pumps; Pumpe D{sub 2}O

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V; Nikolic, M

    1963-12-15

    Continuous increase of radiation intensity was observed on all the elements in the heavy water system during first three years of RA reactor operation. The analysis of heavy water has shown the existence of radioactive cobalt. It was found that cobalt comes from stellite, cobalt based alloy which was used for coating of the heavy water pump discs in order to increase resistance to wearing. Cobalt was removed from the surfaces due to friction, and transferred by heavy water into the reactor where it has been irradiated for 29 876 MWh up to 8-15 Ci/g. Radioactive cobalt contaminated all the surfaces of aluminium and stainless steel parts. This report includes detailed description of heavy water pumps repair, exchange of stellite coated parts, decontamination of the heavy water system, distillation of heavy water. [Serbo-Croat] U toku prve tri godine eksploatacije reaktora RA uocen je neprekidni porast intenziteta zracenja na svim elementima u teskovodnom sistemu. Analizom teske vode utvrdjeno je postojanje radioaktivnog kobalta. Ustanovljeno je da kobalt potice od stelita, legure na bazi kobalta kojim su presvuceni rukavci vratila teskovodnih pumpi radi otpornosi na habanje. Kobalt je trenjem skidan sa povrsina, u toku rada prenosen je teskom vodom u reaktor i ozracivan u toku 29 876 MWh do specificne aktivnosti 8-15 Ci/g. Radioaktivni kobalt je kontaminirao sve povrsine od aluminijuma i nerdjajuceg celika. Ovaj izvestaj sadrzi detaljan opis remonta pumpi, zamene delova teskovodnih pumpi novim delovima bez stelitnog sloja, dekontaminacije teskovodnog sistema, destilacije teske vode.

  5. An efficient hybrid sulfur process using PEM electrolysis with a bayonet decomposition reactor - HTR2008-58207

    International Nuclear Information System (INIS)

    Gorensek, M. B.; Summers, W. A.; Lahoda, E. J.; Bolthrunis, C. O.; Greyvenstein, R.

    2008-01-01

    The Hybrid Sulfur (HyS) Process is being developed to produce hydrogen by water-splitting using heat from advanced nuclear reactors. It has the potential for high efficiency and competitive hydrogen production cost, and has been demonstrated at a laboratory scale. As a two-step process, the HyS is one of the simplest thermochemical cycles. The sulfuric acid decomposition reaction is common to all sulfur cycles, including the Sulfur-Iodine (SI) cycle. What distinguishes the HyS Process from the other sulfur cycles is the use of sulfur dioxide (SO 2 ) to depolarize the anode of a water electrolyzer. The two critical HyS Process components are the SO 2 - depolarized electrolyzer (SDE), and the high-temperature decomposition reactor. A proton exchange membrane (PEM)- type SDE and a silicon carbide bayonet-type high-temperature decomposition reactor are being developed for DOE's Nuclear Hydrogen Initiative (NHI) by Savannah River National Laboratory (SRNL) and by Sandia National Laboratories (SNL), respectively. The ultimate goal of the NHI-sponsored work is to couple the SDE and the reactor in an integrated laboratory scale experiment to prove the technical readiness of the HyS cycle for the NGNP demonstration. This paper describes the flowsheet that is being prepared to combine these two components into a viable process and presents the latest performance projections and economics for a HyS Process coupled to a PBMR heat source. The basic flowsheet for this process has been described elsewhere [4]. It requires an acid concentration section because the SDE product, which is limited to no more than 50% H 2 SO 4 by cell voltage considerations, is too dilute to be fed directly to the bayonet, which needs at least 65% H 2 SO 4 in the feed for acceptable performance. Optimization involved trade-offs between decomposition reaction and acid concentration heat requirements. The PBMR heat source can split its heat output between the decomposition reaction and either steam

  6. The development of fast reactors - Effects on the Swedish system of management of spent fuel; Utveckling av snabba reaktorer - Paaverkan paa det svenska systemet foer hantering av anvaent braensle

    Energy Technology Data Exchange (ETDEWEB)

    Hans Forsstroem, Hans [SKB International AB, Stockholm (Sweden)

    2013-09-15

    Since the start of the nuclear power era studies have been performed of how to utilise the uranium energy resource in the most effective way. Only about one percent of the energy potential of uranium is utilised in the light water reactors of today. To improve the utilization other types of reactors are needed. With fast reactors theoretically 50-100 times more energy can be extracted from the uranium. This will require reprocessing of the uranium and multiple recycling of the plutonium. Plutonium and uranium can also be recycled in light water reactors, but this will only improve the uranium utilisation by about 20 %. Recycling of plutonium on a routine basis is presently only done in France. The development of fast reactors has been going on since the end of the 1940ies. During the 1970ies the planning was that a large number of fast reactors and their associated fuel cycle facilities would be in operation by the year 2000. The development has, however, for different reasons been much slower than planned. The general assessment today is that fast reactor, if they will be realised, will hardly give an important contribution to energy production until after 2050. Nuclear power production has instead been dominated by light water reactors similar to the ones in use in Sweden. Light water reactors are believed to continue to dominate during the next decades. To start a fast reactor system plutonium (or highly enriched uranium) will be needed. Such plutonium is contained in spent nuclear fuel from light water reactors. This raises the question: Should the spent nuclear fuel be stored so that the potential energy resource in the fuel can be used in the future instead of disposing of it as a waste? The answer to this question will depend on when the material will be useful, i.e. when fast reactors have been introduced on a large scale. It will also depend on the demand for plutonium at this time, i.e. will plutonium be a scarce redundant resource at this point of time

  7. RA reactor kinetic parameters - Progress report; Kineticki parametri reaktora RA - Izvestaj o napredovanju -

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, M; Obradovic, D; Jevtovic, V; Velickovic, Lj [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    The objective of nuclear reactor kinetics study is to analyze the stability of reactor operation in practice. The obtained parameters should define the needed properties of automatic control system relevant for the stability of the designed reactor system. Refining the analytical models is done by using the analysis and interpretation of experimental data. Results of measured the reactor response obtained by using the reactor oscillator ROB-1 are explained by using the space independent model of the zero power reactor, by power reactor model with one feedback circuit, and by a complex model. It was assumed that the perturbations of the system are small and that linearized kinetic equations could be used. Linearized kinetic equation of the reactor system are transformed into the frequency region in order to analyze the measured values directly. The objective of this paper is to measure the RA reactor kinetics parameters, and analyze the stability of reactor operation at power levels high than nominal. Istrazivanja u oblasti kinetike nuklearnih reaktora imaju za cilj da dovedu analizu stabilnosti rada reaktora na nivo 'radne tehnologije'. Dobijeni pararametri treba da specificiraju potrebne karakteristike sistema automatske kontrole za odgovarajucu stabilnost projektovanog reaktorskog sistema. Doterivanjem analitickih modela do takvog nivoa da se zapazeni fenomeni mogu anailitcki predvideti ide preko analize i interpretacije eksperimentalnih podataka. Eksperimentalni rezultati merenja odziva reaktora, izvedeni reaktorskim oscilatorom ROB-1, interpretirani su na osnovu prostorno nezavisnog modela za reaktor nulte snage, modelom reaktora snage sa jednim kolom povratne sprege, kao i kompleksnim modelom. U ovom radu se poslo od toga da su perturbacije parametara sistema male, pa se mogu upotrebiti linearizovane kineticke jednacine. Linearizovane kineticke jednacine reaktorskog sistema transformirane su u frekventno podrucje s ciljem direktne analize mernih rezultata