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Sample records for plainsboro irl pool type reactor

  1. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  2. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  3. Radiological performance of hot water layer system in open pool type reactor

    Directory of Open Access Journals (Sweden)

    Amr Abdelhady

    2013-06-01

    Full Text Available The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than the worker permissible dose limits to values very higher than the permissible dose limits.

  4. Design guide for Category III reactors: pool type reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

  5. Declassification of radioactive water from a pool type reactor after nuclear facility dismantling

    Science.gov (United States)

    Arnal, J. M.; Sancho, M.; García-Fayos, B.; Verdú, G.; Serrano, C.; Ruiz-Martínez, J. T.

    2017-09-01

    This work is aimed to the treatment of the radioactive water from a dismantled nuclear facility with an experimental pool type reactor. The main objective of the treatment is to declassify the maximum volume of water and thus decrease the volume of radioactive liquid waste to be managed. In a preliminary stage, simulation of treatment by the combination of reverse osmosis (RO) and evaporation have been performed. Predicted results showed that the combination of membrane and evaporation technologies would result in a volume reduction factor higher than 600. The estimated time to complete the treatment was around 650 h (25-30 days). For different economical and organizational reasons which are explained in this paper, the final treatment of the real waste had to be reduced and only evaporation was applied. The volume reduction factor achieved in the real treatment was around 170, and the time spent for treatment was 194 days.

  6. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

    2015-08-15

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

  7. Feasibility analysis of the Primary Loop of Pool-Type Natural Circulating Nuclear Reactor Dedicated to Seawater Desalination

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woonho; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, the feasibility of natural circulation was evaluated for the reference plant AHR400 (Advanced Heating Reactor 400MWth). AHR400 is a pool-type desalination-dedicated nuclear reactor. As a consequence, AHR400 has low operating pressure and temperature which provides large safety margin. Removal of the reactor coolant pump from the AHR400 will enforce integrity of the reactor vessel and passive safety feature. Therefore, the study also tried to find out optimized primary loop design to achieve total natural circulation of the coolant. Natural circulation capacity of the primary loop of the desalination dedicated nuclear reactor AHR400 was evaluated. It was concluded that to remove RCP from the AHR400 and operates the reactor only by natural circulation of the coolant is impossible. Decreased core power as half make removal of RCP possible with 15m central height difference between the core and IHXs. Furthermore, validation and modification of pressure loss coefficients by small-scaled natural circulation experiment at a pool-type reactor would provide more accurate results.

  8. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  9. Neutron activation analysis at the Livermore pool-type reactor for the environmental research program. [Identification of trace element contaminants

    Energy Technology Data Exchange (ETDEWEB)

    Ragaini, R.C.; Heft, R.E.; Garvis, D.

    1976-07-02

    Instrumental neutron activation analysis is a technique of trace analysis using measurements of radioactivity induced in the sample by exposure to a source of neutrons. The induced activity is measured by the emitted gamma radiation. Each gamma emitter can then be identified by the energy of the photopeaks produced as the nuclide decays and by the half-life of the neutron-induced activity. A complex computer program GAMANAL has been used to accomplish the major tasks of nuclide identification and quantification. The nuclide data output from GAMANAL is processed by a second computer code NADAC, which develops elemental abundance data from disintegration rates observed. The methods are those employed at the Livermore Pool-Type Reactor in support of the environmental research trace element analysis program. Among the procedures described and discussed are sample preparation, irradiation, analysis, and application of the technique.

  10. An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank

    Directory of Open Access Journals (Sweden)

    Kwon-Yeong Lee

    2015-01-01

    Full Text Available In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves.

  11. Successful scaling-up of self-sustained pyrolysis of oil palm biomass under pool-type reactor.

    Science.gov (United States)

    Idris, Juferi; Shirai, Yoshihito; Andou, Yoshito; Mohd Ali, Ahmad Amiruddin; Othman, Mohd Ridzuan; Ibrahim, Izzudin; Yamamoto, Akio; Yasuda, Nobuhiko; Hassan, Mohd Ali

    2016-02-01

    An appropriate technology for waste utilisation, especially for a large amount of abundant pressed-shredded oil palm empty fruit bunch (OFEFB), is important for the oil palm industry. Self-sustained pyrolysis, whereby oil palm biomass was combusted by itself to provide the heat for pyrolysis without an electrical heater, is more preferable owing to its simplicity, ease of operation and low energy requirement. In this study, biochar production under self-sustained pyrolysis of oil palm biomass in the form of oil palm empty fruit bunch was tested in a 3-t large-scale pool-type reactor. During the pyrolysis process, the biomass was loaded layer by layer when the smoke appeared on the top, to minimise the entrance of oxygen. This method had significantly increased the yield of biochar. In our previous report, we have tested on a 30-kg pilot-scale capacity under self-sustained pyrolysis and found that the higher heating value (HHV) obtained was 22.6-24.7 MJ kg(-1) with a 23.5%-25.0% yield. In this scaled-up study, a 3-t large-scale procedure produced HHV of 22.0-24.3 MJ kg(-1) with a 30%-34% yield based on a wet-weight basis. The maximum self-sustained pyrolysis temperature for the large-scale procedure can reach between 600 °C and 700 °C. We concluded that large-scale biochar production under self-sustained pyrolysis was successfully conducted owing to the comparable biochar produced, compared with medium-scale and other studies with an electrical heating element, making it an appropriate technology for waste utilisation, particularly for the oil palm industry.

  12. Stade NPP. Dismantling of the reactor pool

    Energy Technology Data Exchange (ETDEWEB)

    Scharf, Daniel; Dziwis, Joachim [E.ON Anlagenservice GmbH Nukleartechnik, Gelsenkirchen (Germany); Kemp, Lutz-Hagen [KKW Stade GmbH und Co. oHG, Stade (Germany)

    2012-11-01

    Within the scope of the 4{sup th} partial decommissioning permission of Stade NPP the activated and contaminated structures of the reactor pool had to be dismantled in order to gain a completely non-radioactive reactor pool area for the subsequent clearance measurement of the reactor building. In order to achieve the aim it was intended to remove the activated pool liner sheets, its activated framework and several contaminated ventilation channels made of stainless steel, the concrete walls of the reactor pool entirely or in parts depending on their activation level, as well as the remaining activated carbon steel structures of the reactor pool bottom. Embedded in the concrete walls there were several highly contaminated excore tubes and the contaminated pool top edge, which were intended to be removed to its full extent. The contract of the Stade NPP initiated reactor pool dismantling project had been awarded to E.ON Anlagenservice GmbH (EAS) and its subsupplier sat. Kerntechnik GmbH for the concrete dismantling works and was performed as follows. In order to minimize the radiation level in the main working area in accordance with the ALARA principle, the liner sheets and middle parts of its framework were removed by means of angle grinders first, as they were the most dose rate relevant parts. As a result the primary average radiation level in the reactor pool (measured in a distance of 500 mm from the walls) was lowered from 40 {mu}Sv/h to less than 2 {mu}Sv/h. After the minimization of the radiation level in the working area the main dismantling step started with the cutting of the reactor pool walls in blocks by means of diamond rope cutters. Once a concrete block was cut out, it was transported into the fuel pool by means of a crane and crane fork, examined radiologically, marked area by area and segmented to debris by means of an electrical excavator with a hydraulic chisel. Afterwards the debris and carbon steel parts were fractioned and packed for further

  13. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

  14. A model for the analysis of loss of decay heat removal during loss of coolant accident in MTR pool type research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bousbia-salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2, 56126 Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; Meftah, Brahim [Division Reacteur - Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA - Algiers (Algeria); Hamidouche, Tewfik [Laboratoire des Analyses de Surete, Centre de Recherche Nucleaire d' Alger (CRNA), 02 Boulevard Frantz Fanon, B.P. 399, 16000 Algiers (Algeria)]. E-mail: thamidouche@comena-dz.org; Si-Ahmed, El Khider [Laboratoire des Ecoulements Polyhpasiques, Universite des Sciences et de la Technologie d' Alger, Algiers (Algeria)

    2006-03-15

    During a loss of coolant accident leading to total emptying of the reactor pool, the decay heat could be removed through air natural convection. However, under partial pool emptying the core is partially submerged and the coolant circulation inside the fuel element could no more be possible. Under such conditions, a core overheat takes place, and the thermal energy is essentially diffused from the core to its periphery by combined thermal radiation and conduction. In order to predict fuel element temperature evolution under such conditions a mathematical model is performed. The model is based on a 3D geometry and takes into account a variety of core configurations including fuel elements (standard and control), reflector elements and grid plates. The homogeneous flow model is used and the fluid conservation equations are solved using a semi-implicit finite difference method. Preliminary tests of the developed model were made by considering a series of hypothetical accidents. In the current framework a loss of decay heat removal accidents in the IAEA benchmark open pool MTR-type research reactor is considered. It is shown that in the case of a low core immersion height no water boiling is observed and the fuel surface temperature rise remains below the melting point of the aluminium cladding.

  15. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kawakami, Hiroto

    1995-02-07

    A reactor container of the present invention has a structure that the reactor container is entirely at the same temperature as that at the inlet of the reactor and, a hot pool is incorporated therein, and the reactor container has is entirely at the same temperature and has substantially uniform temperature follow-up property transiently. Namely, if the temperature at the inlet of the reactor core changes, the temperature of the entire reactor container changes following this change, but no great temperature gradient is caused in the axial direction and no great heat stresses due to axial temperature distribution is caused. Occurrence of thermal stresses caused by the axial temperature distribution can be suppressed to improve the reliability of the reactor container. In addition, since the laying of the reactor inlet pipelines over the inside of the reactor is eliminated, the reactor container is made compact and the heat shielding structures above the reactor and a protection structure of container walls are simplified. Further, secondary coolants are filled to the outside of the reactor container to simplify the shieldings. The combined effects described above can improve economical property and reliability. (N.H.).

  16. Proposal of novel method of continuous monitoring of possible fuel failure of a pool-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sasaki, K. [Rikkyo University, Nishi-Ikebukuro, Toshima-ku, Tokyo (Japan). College of Science; Hayashi, S.A.; Matsura, T. [Rikkyo University, Nagasaka, Yokosuka (Japan). Institute for Atomic Energy

    1997-10-01

    During the course of studies on fuel failure detection, we have found that the bubbling of a gas such as nitrogen into a reactor coolant water effectively purges the dissolved fission rare gases ({sup 89}Kr, T{sub 1/2}=3.15 min, and {sup 138}Xe, T{sub 1/2}=14.08 min) and that the respective daughter nuclides ({sup 89}Rb, T{sub 1/2}=15.15 min and {sup 138}Cs, T{sub 1/2}=33.41 min) are detected in the washing water of the collected gas mixture. The detected activity depends on the time of standing between sampling and washing of the gas, and the dependence agreed well with the theoretical prediction from the consecutive radioactive decay for both pairs ({sup 89}Kr-{sup 89}Rb, and {sup 138}Xe-{sup 138}Cs). Based on these findings, we have recently constructed a semi-continuous fuel monitoring system, which consists of an automatic and intermittent gas sampler (1 litre bottles) and a bottle conveying unit. After standing for a definite time, bottled gas is shaken with a small amount of water, and the activity of the water is measured. This system operates satisfactorily, but the whole system involves several sophisticated steps so that is rather costly. Quite recently we have got an idea of a simpler, more economical, fully automated continuous system. The system consists in principle only of a large cylinder with packing materials just as in a fractional distiller. On the top of the cylinder there are an inlet of washing water and an outlet of the gas, and at the bottom there are an inlet of the collected gas from the coolant and an outlet of the washing water. The whole system can be operated fully automatically and continuously, with continuous feeding of bubbling gas into the reactor coolant. This has not yet been experimentally tested at present, and in this presentation, information about the setup parameters such as the flow rate of the bubbling gas, the volume of the cylinder and vacant space, the flow rate of the washing water, etc. are reported

  17. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1997-04-04

    An LMFBR type reactor comprises a plurality of reactor cores in a reactor container. Namely, a plurality of pot containing vessels are disposed in the reactor vessel and a plurality of reactor cores are formed in a state where an integrated-type fuel assembly is each inserted to a pot, and a coolant pipeline is connected to each of the pot containing-vessel to cool the reactor core respectively. When fuels are exchanged, the integrated-type fuel assembly is taken out together with the pot from the reactor vessel in a state where the integrated-type fuel assembly is immersed in the coolants in the pot as it is. Accordingly, coolants are supplied to each of the pot containing-vessel connected with the coolant pipeline and circulate while cooling the integrated-type fuel assembly for every pot. Then, when the fuels are exchanged, the integrated type fuel assembly is taken out to the outside of the reactor together with the pot by taking up the pot from the pot-containing vessel. Then, neutron economy is improved to thereby improve reactor power and the breeding ratio. (N.H.)

  18. Development of Pool-type Sodium-cooled Fast Reactor System Analysis Code%池式钠冷快堆系统分析程序开发

    Institute of Scientific and Technical Information of China (English)

    王晋; 张东辉; 胡文军

    2016-01-01

    针对池式钠冷快堆的特点,在对快堆系统的水力模型、热工模型和中子动力学模型进行详细分类和建模的基础上,利用 FORTRAN95语言开发了可用于池式钠冷快堆事故分析的系统分析程序(FASYS程序)。以中国实验快堆为计算对象对FASYS程序模型进行了初步验证,所获得的结果和试验值与其他系统程序计算值符合良好,证明了所开发的系统分析程序的正确性。%According to the characteristics of pool‐type sodium‐cooled fast reactor ,and with the fast reactor hydraulic model , thermal model and neutron kinetics model thoroughly classified and developed ,a fast reactor system analysis code (FASYS code) was developed by FORTRAN95 language for pool‐type sodium‐cooled fast reactor acci‐dent analysis .Transient conditions in CEFR were calculated with FASYS code and the results were used for code validation .The calculation results are consistent with the test data and other fast reactor system analysis code results , and the correctness of the FASYS code is proved .

  19. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, Takeshi; Iida, Masaaki; Moriki, Yasuyuki

    1994-10-18

    A reactor core is divided into a plurality of coolants flowrate regions, and electromagnetic pumps exclusively used for each of the flowrate regions are disposed to distribute coolants flowrates in the reactor core. Further, the flowrate of each of the electromagnetic pumps is automatically controlled depending on signals from a temperature detector disposed at the exit of the reactor core, so that the flowrate of the region can be controlled optimally depending on the burning of reactor core fuels. Then, the electromagnetic pumps disposed for every divided region are controlled respectively, so that the coolants flowrate distribution suitable to each of the regions can be attained. Margin for fuel design is decreased, fuels are used effectively, as well as an operation efficiency can be improved. Moreover, since the electromagnetic pump has less flow resistance compared with a mechanical type pump, and flow resistance of the reactor core flowrate control mechanism is eliminated, greater circulating flowrate can be ensured after occurrence of accident in a natural convection using a buoyancy of coolants utilizable for after-heat removal as a driving force. (N.H.).

  20. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  1. 用于池式快堆系统分析的钠池三维模型开发%Development of Three-Dimensional Sodium Pool Model for System Analysis of Pool-Type Liquid Metal Fast Breeder Reactor

    Institute of Scientific and Technical Information of China (English)

    隋丹婷; 陆道纲; 张盼

    2012-01-01

    由于池式快堆钠池内的热工水力学特性对反应堆的安全运行有重要影响,本文采用基于交错网格的SIMPLE算法开发直角坐标系和柱坐标系下钠池三维计算软件.应用CFX软件进行验证之后,完成了三维流场分析程序与系统分析软件SAC-CFR的耦合,并用耦合后的程序分析日本文殊快堆45%功率稳态运行工况上腔室内的流场分布,初步验证了堆芯上腔三维化的SAC-CFR用于系统分析的有效性,为进一步开发事故模型、非能动余热排出系统模型做准备.%As the thermal-hydraulic characteristic in sodium pool is crucial for safety operation of liquid metal fast breeder reactor (LMFBR), a three-dimensional sodium pool thermal-hydraulic analysis code was developed based on SIMPLE algorithm on stagger grid under Cartesian coordinates and cylindrical coordinates. After the validation with CFX, coupling between the analysis code and SAC-CFR was completed) and then the coupled code was applied to the flow field analysis in upper plenum of Monju Plant at 45% thermal power steady-state operation condition, which preliminary shows the effectiveness of the system analysis with coupled code and makes preparations for further development of accident analysis model and passive residual heat removal system.

  2. Häälekas IRL / Andrus Saar

    Index Scriptorium Estoniae

    Saar, Andrus, 1946-

    2011-01-01

    Autor vaatleb, miks IRL ei suuda tõusta usaldusreitingute järgi teiseks suureks erakonnaks. Autori hinnangul ei ole Isamaaliidu ja Res Publica pragmaatilisest kooselust tekkinud ühtset selgelt eristuvat poliitilist nägu, millega suurem osa valijaskonnast saaks end usaldusväärselt identifitseerida

  3. Decommissioning of the pool reactor Thetis in Ghent, Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Cortenbosch, Geert; Mommaert, Chantal [Bel V, Brussels (Belgium); Tierens, Hubert; Monsieurs, Myriam; Meierlaen, Isabelle; Strijckmans, Karel [Ghent Univ. (Belgium)

    2016-11-15

    The Thetis research pool reactor (with a nominal power of 150 kW) of the Ghent University was operational from 1967 till December 2003. The first phase of the decommissioning of the reactor, the removal of the spent fuel from the site, took place in 2010. The cumulative dose received was only 404 man . μSv. During the second phase, the transition period between the removal of the spent fuel in 2010 and the start of the decommissioning phase in March 2013, 3-monthly internal inspections and inspections by Bel V, were performed. The third and final decommissioning phase started on March 18, 2013. The total dose received between March 2013 and August 2013 was 1561 man . μSv. The declassification from a Class I installation to a Class II installation was possible by the end of 2015. The activated concrete in the reactor pool will remain under regulatory control until the activation levels are lower than the limits for free release.

  4. LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iwashige, Kengo

    1996-06-21

    In an LMFBR type reactor, partitions are disposed to a coolant channel at positions lower than the free liquid level, and the width of the partitions is adapted to have a predetermined condition. Namely, when low temperature fluid overflowing the wall of the coolant channel, flows down and collided against the free liquid surface in the coolant channel, since the dropping speed thereof is reduced abruptly, large pressure waves are caused by kinetic force of the low temperature fluid. However, if appropriate numbers of partitions having an appropriate shape are formed, the dropping speed of the low temperature fluid is moderated to reduce the pressure waves. In addition, since the pressure waves are dispersed to the circumferential and lateral directions of the coolant flow channel respectively, the propagation of the pressure waves can be prevented effectively. Further, when the flow of the low temperature fluid is changed to the circumferential direction, for example, by earthquakes, since the partitions act as members resisting against the circumferential change of the low temperature fluid, the change of the direction can be suppressed. (N.H.)

  5. New reactor type proposed

    CERN Multimedia

    2003-01-01

    "Russian scientists at the Research Institute of Nuclear Power Engineering in Moscow are hoping to develop a new reactor that will use lead and bismuth as fuel instead of uranium and plutonium" (1/2 page).

  6. Measurement of the Residual Stresses and Investigation of Their Effects on a Hardfaced Grid Plate due to Thermal Cycling in a Pool Type Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    S. Balaguru

    2016-01-01

    Full Text Available In sodium-cooled fast reactors (SFR, grid plate is a critical component which is made of 316 L(N SS. It is supported on core support structure. The grid plate supports the core subassemblies and maintains their verticality. Most of the components of SFR are made of 316 L(N/304 L(N SS and they are in contact with the liquid-metal sodium which acts as a coolant. The peak operating temperature in SFR is 550°C. However, the self-welding starts at 500°C. To avoid self-welding and galling, hardfacing of the grid plate has become necessary. Nickel based cobalt-free colmonoy 5 has been identified as the hardfacing material due to its lower dose rate by Plasma Transferred Arc Welding (PTAW. This paper is concerned with the measurement and investigations of the effects of the residual stress generated due to thermal cycling on a scale-down physical model of the grid plate. Finite element analysis of the hardfaced grid plate model is performed for obtaining residual stresses using elastoplastic analysis and hence the results are validated. The effects of the residual stresses due to thermal cycling on the hardfaced grid plate model are studied.

  7. Development of System Analysis Code for Pool-Type Fast Reactor Under Transient Operation%池式快堆系统瞬态分析软件开发

    Institute of Scientific and Technical Information of China (English)

    陆道纲; 隋丹婷

    2012-01-01

    为实现快堆系统分析软件国产化,在已开发的适用于稳态计算的池式快堆系统分析软件SAC-CFR的基础上,进一步开发了系统各部件的瞬态模型、控制系统和保护系统模型、瞬态工况热工水力学的求解逻辑,完成瞬态计算功能的开发.通过对日本文殊快堆45%功率汽机跳闸工况进行建模分析,验证了SAC-CFR用于系统瞬态分析的有效性,为进一步开发非能动余热排出系统分析模型打下了基础.%Aiming at developing system analysis code independently, the system analysis code for pool-type fast reactor in China (SAC-CFR) under transient operation was developed with further development of component transient model, plant control and protection system model, calculation logic for system transient thermal-hydraulic analysis based on the former SAC-CFR version applicable to steady state analysis. The transient started from turbine trip test at 45 % thermal output in the Monju Plant was analyzed with the developed SAC-CFR. A good agreement between the calculated results and the test data was obtained. SAC-CFR is now ready to incorporate passive residual heat removal model for China Experimental Fast Reactor.

  8. Overview of pool hydraulic design of Indian prototype fast breeder reactor

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  9. Natural and mixed convection in the cylindrical pool of TRIGA reactor

    Science.gov (United States)

    Henry, R.; Tiselj, I.; Matkovič, M.

    2017-02-01

    Temperature fields within the pool of the JSI TRIGA MARK II nuclear research reactor were measured to collect data for validation of the thermal hydraulics computational model of the reactor tank. In this context temperature of the coolant was measured simultaneously at sixty different positions within the pool during steady state operation and two transients. The obtained data revealed local peculiarities of the cooling water dynamics inside the pool and were used to estimate the coolant bulk velocity above the reactor core. Mixed natural and forced convection in the pool were simulated with a Computational Fluid Dynamics code. A relatively simple CFD model based on Unsteady RANS turbulence model was found to be sufficient for accurate prediction of the temperature fields in the pool during the reactor operation. Our results show that the simple geometry of the TRIGA pool reactor makes it a suitable candidate for a simple natural circulation benchmark in cylindrical geometry.

  10. 多孔介质方法在池式快堆系统分析软件SAC-CFR三维钠池计算模型中的应用%Application of Porous Medium Method on Three-Dimensional Sodium Pool Model for Pool-Type Fast Reactor System Analysis Code SAC-CFR

    Institute of Scientific and Technical Information of China (English)

    隋丹婷; 陆道纲; 任丽霞; 刘一哲

    2012-01-01

    为准确分析池式快堆热钠池内的热工水力学特性,在已开发出的用于池式快堆系统分析的钠池三维计算模型的基础上,应用多孔介质方法建立钠池内中间热交换器、主泵、事故热交换器及屏蔽柱模型,完成了含有多孔介质模型和复杂几何边界的钠池三维计算模型开发.将该模型嵌入池式快堆系统分析软件SAC-CFR后,分析了中国实验快堆在稳态运行和紧急停堆工况下钠池内的流场分布,初步证明了所采用的多孔介质模型的合理性,为下一步非能动余热排出系统模型的开发做准备.%To simulate the fluid dynamic and thermal characteristics in sodium pool accurately , newly three-dimensional hot pool analysis model with porous medium model and complex geometry was developed after incorporating the porous model of penetration components into three-dimensional model developed already for system analysis of pool-type fast reactor. Penetration components include intermediate heat exchanger, primary pump, decay heat exchanger, and radial shielding. After coupling with the system analysis code SAC-CFR, the newly coupled code was applied to analyze the flow field in hot pool under steady-state operation condition and after scram. The agreement between the computational flow field and the geometry of hot pool shows the effectiveness of porous medium model, which makes preparations for further development of passive residual heat removal system.

  11. Preliminary Requirement of Hot Pool Free Surface Level from PGSFR Reactor Head

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeonghoi; Joo, Hyeongkook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The sensitivity study on structural integrity evaluations are carried out to make a decision of a hot pool free surface location from the reactor head for a preliminary designed reactor enclosure system. To do this, the thermal stress evaluations for a reactor vessel are carried out for a steady state normal operating condition with detailed heat transfer analyses through the reactor enclosure system. From these results, the preliminary design requirement of a hot pool free surface location from the reactor head is established to be 2.0m. From the sensitivity studies on the structural integrity evaluations for a steady state condition, the preliminary distance from the hot pool free surface to the reactor head is determined to be 2.0m same as a conceptual design. More detailed structural analyses for a reactor enclosure system will be carried out as a PGSFR structural design goes forward in detail.

  12. Gas entrainment in scaled model of pool type LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Banerjee, I.; Chandra, L.; Laxman, D.; Kumar, A.; Gopal, C.A.; Shivakumar, N.S.; Padmakumar, G.; Anand Babu, C.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)

    2007-07-01

    The reactor Thermal hydraulics plays an important role for successful operation of Prototype Fast Breeder Reactor (PFBR), which is under construction at Kalpakkam, India. One of the issues to be resolved in PFBR is argon cover gas entrainment problem from free liquid sodium surface. The entrained cover gas may hinder the normal reactor operation. High free surface velocity along with the presence of various immersed components in the hot pool is the cause of gas entrainment from free surface. To reduce the free surface velocity and hence gas entrainment, ring type baffle plates were considered. Initially the optimum geometry of the baffle plate was arrived through numerical analysis using PHOENICS, a commercial computational fluid dynamics tool. Finally the experiments were conducted in a 1/4 scale water model of PFBR primary circuit with selected baffle plate geometry. It was found that a baffle plate with radial width of 125 mm in the model and located above intermediate heat exchanger is very effective to reduce the gas entrainment problem in PFBR. (authors)

  13. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  14. Fast Reactor Fuel Type and Reactor Safety Performance

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; J. Cahalan

    2009-09-01

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and

  15. Summary of Research on Interactive Safety Analysis Program of Pool Type SFR

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    The aim of the project was to develop an interactive safety analysis program of pool type sodium cooled fast reactors (SFR), based on a French system code OASIS. The function and physical model of the program should be verified by the application on CEFR design.

  16. meIRL-BC: Predicting Player Positions in Video Games

    NARCIS (Netherlands)

    Becht, I.; Bakkes, S.; Barnes, T.; Bogost, I.

    2014-01-01

    In this paper we demonstrate how behaviour-classification models can improve player position prediction for video game AI. To this end, we propose a novel method named meIRL-BC, which (1) uses maximum-entropy Inverse Reinforcement Learning for the creation of position prediction models [15], and (2)

  17. Criticality safety calculations of the Soreq research reactor storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Caner, M.; Hirshfeld, H.; Nagler, A.; Silverman, I.; Bettan, M. [Soreq Nuclear Research Center, Yavne 81800 (Israel); Levine, S.H. [Penn State University, University Park 16802 (United States)

    2001-07-01

    The IRR-l spent fuel is to be relocated in a storage pool. The present paper describes the actual facility and summarizes the Monte Carlo criticality safety calculations. The fuel elements are to be placed inside cadmium boxes to reduce their reactivity. The fuel element is 7.6 cm by 8.0 cm in the horizontal plane. The cadmium box is effectively 9.7 cm by 9.7 cm, providing significant water between the cadmium and the fuel element. The present calculations show that the spent fuel storage pool is criticality safe even for fresh fuel elements. (author)

  18. Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

    Energy Technology Data Exchange (ETDEWEB)

    Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-05-15

    Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

  19. Feynman-alpha technique for measurement of detector dead time using a 30 kW tank-in-pool research reactor

    CERN Document Server

    Akaho, E H K; Intsiful, J D K; Maakuu, B T; Nyarko, B J B

    2002-01-01

    Reactor noise analysis was carried out for Ghana Research Reactor-1 GHARR-1, a tank-in-pool type reactor using the Feynman-alpha technique (variance-to-mean method). Measurements made at different detector positions and under subcritical conditions showed that the technique could not be used to determine the prompt decay constant for the reactor which is Be reflected with photo-neutron background. However, for very low dwell times the technique was used to measure the dead time of the detector which compares favourably with the value obtained using the alpha-conventional method.

  20. Dynamic simulation of accidental closure of intermediate heat exchanger isolation valve in a pool type LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K., E-mail: natesan@igcar.gov.in [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Kasinathan, N.; Velusamy, K.; Selvaraj, P.; Chellapandi, P.; Chetal, S.C. [Reactor Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2011-04-15

    Research highlights: > Thermal hydraulic analysis closure of sleeve valve in the primary circuit of FBR is discussed. > Numerical modeling of hydraulics in the primary and secondary sodium circuits is presented. > Aspects related to event management are discussed. - Abstract: In a pool type liquid metal cooled fast breeder reactor (LMFBR), core and other internals such as pumps, heat exchangers are immersed in a pool of sodium. Heat exchange from primary sodium circuit (pool) to secondary sodium circuit (loop) is through four intermediate heat exchangers (IHX) immersed in primary sodium pool. Each IHX is provided with a sleeve valve at its primary sodium inlet window for the purpose of isolating the shell side of IHX from the sodium pool. With such a provision, an inadvertent partial or complete closure of a sleeve valve of one of the IHX during normal operation of the reactor has been considered as a design basis event for the reactor. One dimensional transient thermal hydraulic models of the primary and secondary sodium circuits have been developed to study the thermal hydraulic consequences of such an event. The main areas of concern in the plant and the availability of safety parameters for the detection of the event have been evaluated.

  1. Aaviksoo soovib IRL-ile leida uue peasekretäri / Mirko Ojakivi

    Index Scriptorium Estoniae

    Ojakivi, Mirko

    2010-01-01

    IRL-i poliitik, kaitseminister Jaak Aaviksoo on seisukohal, et erakonna organisatsioonilise juhtimisega peaks tegelema isik, kes pole iga päev seotud muu tööga. IRL-i peasekretär Margus Tsahkna on samas ka Riigikogu liige. IRL-i esimeheks kandideerib Mart Laar, aseesimeesteks Jaak Aaviksoo, Ene Ergma, Tõnis Lukas ja Juhan Parts

  2. Cost estimates of operating onsite spent fuel pools after final reactor shutdown

    Energy Technology Data Exchange (ETDEWEB)

    Rod, S R

    1991-08-01

    This report presents estimates of the annual costs of operating spent fuel pools at nuclear power stations after the final shutdown of one or more onsite reactors. Its purpose is to provide basic spent fuel storage cost information for use in evaluating DOE's reference nuclear waste management system, as well as alternate systems. The basic model of an independent spent fuel storage installation (ISFSI) used in this study was based on General Electric Corporation's Morris Operation and was modified to reflect mean storage capabilities at an unspecified, or generic,'' US reactor site. Cost data for the model came from several sources, including both operating and shutdown nuclear power stations and existing ISFSIs. Duke Power Company has estimated ISFSI costs based on existing spent fuel storage costs at its nuclear power stations. Similarly, nuclear material handling facilities such as the Morris Operation, the West Valley Demonstration Project, and the retired Humbolt Bay nuclear power station have compiled spent fuel storage cost data based on years of operating experience. Consideration was given to the following factors that would cause operating costs to vary among pools: (1) The number of spent fuel pools at a given reactor site; (2) the number of operating and shutdown reactors onsite; (3) geographic location; and (4) pool storage capacity. 10 ref., 6 figs., 7 tabs.

  3. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  4. 3-dimensional thermohydraulic analysis of KALIMER reactor pool during unprotected accidents

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Hahn Do Hee

    2003-01-01

    During a normal reactor scram, the heat generation is reduced almost instantaneously while the coolant flow rate follows the pump coastdown. This mismatch between power and flow results in a situation where the core flow entering the hot pool is at a lower temperature than the temperature of the bulk pool sodium. This temperature difference leads to thermal stratification. Thermal stratification can occur in the hot pool region if the entering coolant is colder than the existing hot pool coolant and the flow momentum is not large enough to overcome the negative buoyancy force. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response. Hence, it is necessary to predict the pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs and its contribution to the net buoyancy head. Therefore, two-dimensional hot pool thermohydraulic model named HP2D has been developed. In this report code-to-code comparison analysis between HP2D and COMMIX-1AR/P has been performed in the case of steady-state and UTOP.

  5. Advanced thermohydraulic simulation code for pool-type LMFBRs (SSC-P code)

    Energy Technology Data Exchange (ETDEWEB)

    Madni, I.K.; Cazzoli, E.G.

    1980-09-01

    Models for components and processes that are needed for simulation of thermohydraulic transient in a pool-type liquid metal fast breeder reactor (LMFBR) plant are described in this report. A computer code, SSC-P, has been developed as a part of the Super System Code (SSC) development project. A user's manual is being prepared as a separate document. 27 refs., 26 figs., 1 tab.

  6. Socioeconomic information, Plainsboro area, New Jersey: Supplementary documentation for an environmental assessment for the CIT (Compact Ignition Tokamak) at PPPL

    Energy Technology Data Exchange (ETDEWEB)

    Bentz, L.K.; Bender, D.S.

    1987-07-01

    This report contains socioeconomic information on the Plainsboro, New Jersey, area, the proposed location of the Compact Ignition Tokamak (CIT) facility. It was prepared as supplemental information for an environmental assessment for the CIT at Princeton Plasma Physics Laboratory (PPPL). The report contains descriptions of the demographic, economic, and community resource characteristics, and, based on information available in early 1987, considers the socioeconomic effect of the proposed facility. In all areas examined, the anticipated socioeconomic impacts of the proposed CIT facility at PPPL are negligible or minimal. 29 refs., 8 figs., 24 tabs.

  7. Fuel Burnup and Fuel Pool Shielding Analysis for Bushehr Nuclear Reactor VVER-1000

    Science.gov (United States)

    Hadad, Kamal; Ayobian, Navid

    Bushehr Nuclear power plant (BNPP) is currently under construction. The VVER-1000 reactor will be loaded with 126 tons of about 4% enriched fuel having 3-years life cycle. The spent fuel (SF) will be transferred into the spent fuel pool (SPF), where it stays for 8 years before being transferred to Russia. The SPF plays a crucial role during 8 years when the SP resides in there. This paper investigates the shielding of this structure as it is designed to shield the SF radiation. In this study, the SF isotope inventory, for different cycles and with different burnups, was calculated using WIMS/4D transport code. Using MCNP4C nuclear code, the intensity of γ rays was obtained in different layers of SFP shields. These layers include the water above fuel assemblies (FA) in pool, concrete wall of the pool and water laid above transferring fuels. Results show that γ rays leakage from the shield in the mentioned layers are in agreement with the plant's PSAR data. Finally we analyzed an accident were the water height above the FA in the pool drops to 47 cm. In this case it was observed that exposure dose above pool, 10 and 30 days from the accident, are still high and in the levels of 1000 and 758 R/hr.

  8. Moving hot cell for LMFBR type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru

    1994-09-16

    A moving hot cell for an LMFBR type reactor is made movable on a reactor operation floor between a position just above the reactor container and a position retreated therefrom. Further, it comprises an overhung portion which can incorporate a spent fuel just thereunder, and a crane for moving a fuel assembly between a spent fuel cask and a reactor container. Further, an opening/closing means having a shielding structure is disposed to the bottom portion and the overhung portion thereof, to provide a sealing structure, in which only the receiving port for the spent fuel cask faces to the inner side, and the cask itself is disposed at the outside. Upon exchange of fuels, the movable hot cell is placed just above the reactor to take out the spent fuels, so that a region contaminated with primary sodium is limited within the hot cell. On the other hand, upon maintenance and repair for equipments, the hot cell is moved, thereby enabling to provide a not contaminated reactor operation floor. (N.H.).

  9. Detectability prediction for a thermoacoustic sensor in the breazeale nuclear reactor pool

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James [Idaho National Laboratory, Idaho Falls, ID (United States); Hrisko, Joshua [Idaho National Laboratory, Idaho Falls, ID (United States); Garrett, Steven [Idaho National Laboratory, Idaho Falls, ID (United States)

    2016-03-01

    Laboratory experiments have suggested that thermoacoustic engines can be in- corporated within nuclear fuel rods. Such engines would radiate sounds that could be used to measure and acoustically-telemeter information about the op- eration of the nuclear reactor (e.g., coolant temperature or uxes of neutrons or other energetic particles) or the physical condition of the nuclear fuel itself (e.g., changes in temperature, evolved gases) that are encoded as the frequency and/or amplitude of the radiated sound [IEEE Measurement and Instrumen- tation 16(3), 18-25 (2013)]. For such acoustic information to be detectable, it is important to characterize the vibroacoustical environments within reactors. Measurements will be presented of the background noise spectra (with and with- out coolant pumps) and reverberation times within the 70,000 gallon pool that cools and shields the fuel in the 1 MW research reactor on Penn State's campus using two hydrophones, a piezoelectric projector, and an accelerometer. Sev- eral signal-processing techniques will be demonstrated to enhance the measured results. Background vibrational measurement were also taken at the 250 MW Advanced Test Reactor, located at the Idaho National Laboratory, using ac- celerometers mounted outside the reactor's pressure vessel and on plumbing will also be presented. The detectability predictions made in the thesis were validated in September 2015 using a nuclear ssion-heated thermoacoustic sensor that was placed in the core of the Breazeale Nuclear Reactor on Penn State's campus. Some features of the thermoacoustic device used in that experiment will also be revealed. [Work supported by the U.S. Department of Energy.

  10. Inspection of state of spent fuel elements stored in RA reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Bulkin, S.Yu.; Sokolov, A.V. [Research and Development Institute of Power Engineering, Moscow (Russian Federation); Matausek, M.V.; Vukadin, Z. [VINCA Institute of Nuclear Science, Belgrade (Yugoslavia)

    1999-07-01

    About five thousand spent fuel elements from RA reactor have been stored for over 30 years in sealed aluminum barrels in the spent fuel storage pool. This way of storage does not provide complete information about the state of spent fuel elements or the medium inside the barrels, like pressure or radioactivity. The technology has recently been developed and the equipment has been manufactured to inspect the state of the spent fuel and to reduce eventual internal pressure inside the aluminum barrels. Based on the results of this inspection, a procedure will be proposed for transferring spent fuel to a more reliable storage facility. (author)

  11. Review of fuel assembly and pool thermal hydraulics for fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roelofs, Ferry, E-mail: roelofs@nrg.eu; Gopala, Vinay R.; Jayaraju, Santhosh; Shams, Afaque; Komen, Ed

    2013-12-15

    Highlights: • Literature review of fuel assembly and pool thermal hydraulics for fast reactors. • Experiments and state-of-the-art simulations. • For wire wrapped fuel assemblies RANS for complete fuel assembly is state-of-the-art, LES serves reference. • For pool thermal hydraulics, typically 5 to 20 million computational volumes are used in RANS simulations. • Gas entrainment analyses are extremely demanding as in addition they request multiphase modelling. -- Abstract: Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possible efficient use of uranium and the possibility to reduce the volume and lifetime of nuclear waste. Thermal-hydraulics is recognized as a key scientific subject in the development of such reactors. Two important challenges for the design of liquid metal fast reactors (LMFRs) are fuel assembly and pool thermal hydraulics. The heart of every nuclear reactor is the core, where the nuclear chain reaction takes place. Heat is produced in the nuclear fuel and transported to the coolant. LMFR core designs consist of many fuel assemblies which in turn consist of a large number of fuel rods. Wire wraps are commonly envisaged as spacer design in LMFR fuel assemblies. For the design and safety analyses of such reactors, simulations of the heat transport within the core are essential. The flow exiting the core is made up of the outlets of many different fuel assemblies. The liquid metal in these assemblies may be heated up to different temperatures. This leads to temperature fluctuations on various above core structures. As these temperature fluctuations may lead to thermal fatigue damage of the structures, an accurate characterization of the liquid metal flow field in the above core region is very important. This paper will provide an overview of state-of-the-art evaluations of fuel assembly and pool thermal hydraulics for LMFRs. It will show the tight interaction

  12. Thermal hydraulics in the hot pool of Fast Breeder Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Padmakumar, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)], E-mail: gpk@igcar.gov.in; Pandey, G.K.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, TN 603 102 (India)

    2009-06-15

    Sodium cooled Fast Breeder Test Reactor (FBTR) of 40 MWt/13 MWe capacity is in operation at Kalpakkam, near Chennai. Presently it is operating with a core of 10.5 MWt. Knowledge of temperatures and flow pattern in the hot pool of FBTR is essential to assess the thermal stresses in the hot pool. While theoretical analysis of the hot pool has been conducted by a three-dimensional code to access the temperature profile, it involves tuning due to complex geometry, thermal stresses and vibration. With this in view, an experimental model was fabricated in 1/4 scale using acrylic material and tests were conducted in water. Initially hydraulic studies were conducted with ambient water maintaining Froude number similarity. After that thermal studies were conducted using hot and cold water maintaining Richardson similitude. In both cases Euler similarity was also maintained. Studies were conducted simulating both low and full power operating conditions. This paper discusses the model simulation, similarity criteria, the various thermal hydraulic studies that were carried out, the results obtained and the comparison with the prototype measurements.

  13. Reformil valitsuses kaks uut nägu. IRL veel arutab / Alyona Stadnik

    Index Scriptorium Estoniae

    Stadnik, Alyona

    2011-01-01

    Reformierakonnal on ministrikandidaadid paigas ja 26. märstil otsustatakse, kas kinnitada koalitsioonileping Isamaa ja Res Publica Liiduga. IRL avalikustab oma ministrikandidaadid pärast erakonna volikogu istungit

  14. Study on the Adaptability of Etheriifcation Feedstock to Reactor Type

    Institute of Scientific and Technical Information of China (English)

    Mao Junyi; Yuan Qing; Wang Lei; Huang Tao

    2016-01-01

    A reactive C5 oleifns and methanol etheriifcation kinetic model based on E-R mechanism was established and three different types of reactors including the adiabatic ifxed-bed liquid reactor, the external loop reactor and the mixed-phase reactor were constructed by Aspen Plus. The adaptability of reactive C5 oleifns to these reactors was studied and simulated using various gasoline fractions with different oleifns content. After the theoretical model was validated by the experimental data of the etheriifcation of three C5 light cut fractions from different gasoline sources in different reactors, the simulated isoamylene conversion with reactive C5 olefin contents increasing from 10% to 60% was studied in the three different types of reactors for etheriifcation with methanol, respectively. Test results show that there is an obvious adaptability of the feedstock composition to the reactor type to achieve a high conversion.

  15. The radionuclides of primary coolant in HANARO and the recent activities performed to reduce the radioactivity or reactor pool water

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin [HANARO Research Reactor Centre, Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1998-10-01

    In HANARO reactor, there have been activities to identify the principal radionuclides and to quantify them under the normal operation. The purposes of such activities were to establish the measure by which we can reduce the radioactivity of the reactor pool water and detect, in early stage, the abnormal symptoms due to the leakage of radioactive materials from the irradiation sample or the damage of the nuclear fuel, etc. The typical radionuclides produced by the activation of reactor coolant are N{sup 16} and Ar{sup 41}. The radionuclides produced by the activation of the core structural material consist of Na{sup 24}, Mn{sup 56}, and W{sup 187}. Of the various radionuclides, governing the radiation level at the pool surface are Na{sup 24}, Ar{sup 41}, Mn{sup 58}, and W{sup 187}. By establishing the hot water layer system on the pool surface, we expected that the radionuclides such as Ar{sup 41} and Mn{sup 56} whose half-life are relatively short could be removed to a certain extent. Since the content of radioactivity of Na{sup 24} occupies about 60% of the total radioactivity, we assumed that the total radiation level would be greatly reduced if we could decrease the radiation level of Na{sup 24}. However the actual radiation level has not been reduced as much as we expected. Therefore, some experiments have been carried out to find the actual causes afterwards. What we learned through the experiments are that any disturbance in reactor pool water layer causes increase of the pool surface radiation level and even if we maintain the hot water layer well, reactor shutdown will be very much likely to happen once the hot water layer is disturbed. (author)

  16. Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool.

    Science.gov (United States)

    Huang, Chun-Ping; Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin

    2012-09-30

    There were approximately 926 m(3) of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as (137)Cs, (90)Sr, U, and α-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb β-emitters, α-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously eliminate particles and adsorb ionic radionuclides from water.

  17. Experimental validation of thermal design of top shield for a pool type SFR

    Energy Technology Data Exchange (ETDEWEB)

    Aithal, Sriramachandra, E-mail: saithal@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Babu, V. Rajan; Balasubramaniyan, V.; Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Chellapandi, P. [Bharatiya Nabhikiya Vidyut Nigam Limited, Kalpakkam 603102 (India)

    2016-04-15

    Highlights: • Overall thermal design of top shield in a SFR is experimentally verified. • Air jet cooling is effective in ensuring the temperatures limits for top shield. • Convection patterns in narrow annulus are in line with published CFD results. • Wire mesh insulation ensures gradual thermal gradient at top portion of main vessel. • Under loss of cooling scenario, sufficient time is available for corrective action. - Abstract: An Integrated Top Shield Test Facility towards validation of thermal design of top shield for a pool type SFR has been conceived, constructed & commissioned. Detailed experiments were performed in this experimental facility having full-scale features. Steady state temperature distribution within the facility is measured for various heater plate temperatures in addition to simulating different operating states of the reactor. Following are the important observations (i) jet cooling system is effective in regulating the roof slab bottom plate temperature and thermal gradient across roof slab simulating normal operation of reactor, (ii) wire mesh insulation provided in roof slab-main vessel annulus is effective in obtaining gradual thermal gradient along main vessel top portion and inhibiting the setting up of cellular convection within annulus and (iii) cellular convection with four distinct convective cells sets in the annular gap between roof slab and small rotatable plug measuring ∼ϕ4 m in diameter & gap width varying from 16 mm to 30 mm. Repeatability of results is also ensured during all the above tests. The results presented in this paper is expected to provide reference data for validation of thermal hydraulic models in addition to serving as design validation of jet cooling system for pool type SFR.

  18. The noncondensable gas effects on loss-of-coolant accident steam condensation loads in boiling water reactor pressure suppression pool

    Energy Technology Data Exchange (ETDEWEB)

    Kukita, Y.; Namatame, K.; Shiba, M.; Takeshita, I.

    1983-11-01

    The noncondensable gas effects on the loss-ofcoolant-accident-induced steam condensation loads in the boiling water reactor pressure suppression pool have been investigated with regard to experimental data obtained from a large-scale multivent test program. Previous studies have noted that the presence of the noncondensable gas (air), which initially fills the containment drywell space, stabilizes the direct-contact condensation in the pressure suppression pool and hampers onset of the chugging phenomenon, which induces most significant steam condensation load onto the pool boundary. This was found to be true for the tests with relatively small-break diameters, where the maximum steam mass fluxes in the vent pipe were lower than the upper threshold value for the onset of chugging. However, in the tests with the maximum vent steam mass fluxes moderately higher than the chugging upper threshold value, early depletion of the noncondensable gas tended to result in significant stabilization of steam condensation accompanied by an excursion of temperature of pool water surrounding the vent pipe outlets, which led to a delayed onset of chugging. Due to this combined influence of the noncondensable gas and nonuniform pool temperature, and due to dependence of magnitude of chugging load on the vent steam mass flux, the peak magnitude of the steam condensation load appearing in a blowdown can be very sensitive to the initial and break conditions.

  19. Types, Evolution and Pool-Controlling Significance of Pool Fluid Sources in Superimposed Basins: A Case Study from Paleozoic and Mesozoic in South China

    Institute of Scientific and Technical Information of China (English)

    Xu Sihuang; Mei Lianfu; Yuan Caiping; Ma Yongsheng; Guo Tonglou

    2007-01-01

    Having multiple tectonic evolution stages, South China belongs to a superimposed basin in nature. Most marine gas pools became secondary pools. The pool fluid sources serve as the principal pool-controlling factors. On the basis of eight typical petroleum pools, the type, evolution in time-space,and the controlling of petroleum distribution of pool fluid sources are comprehensively analyzed. The main types of pool fluid sources include hydrocarbon, generated primarily and secondly from source rocks, gas cracked from crude oil, gas dissolved in water, inorganic gas, and mixed gases. In terms of evolution, the primary hydrocarbon was predominant prior to Indosinian; during Indosinian to Yenshanian the secondary gas includes gas cracked from crude oil, gas generated secondarily, gas dissolved in water, and inorganic gas dominated; during Yenshanian to Himalayan the most fluid sources were mixed gases. Controlled by pool fluid sources, the pools with mixed gas sources distributed mainly in Upper Yangtze block, especially Sichuan (四川) basin; the pools with primary hydrocarbon sources distributed in paleo-uplifts such as Jiangnan (江南), but most of these pools became fossil pools; the pools with secondary hydrocarbon source distributed in the areas covered by Cretaceous and Eogene in Middle-Lower Yangtze blocks, and Chuxiong (楚雄), Shiwandashan (十万大山), and Nanpanjiang (南盘江) basins; the pools with inorganic gas source mainly formed and distributed in tensional structure areas.

  20. An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, M.; Farmer, M.; Grabaskas, D.

    2017-06-26

    The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

  1. 10 MW research reactor simulation using fuel plate type

    Energy Technology Data Exchange (ETDEWEB)

    Mustafa, M. El Sayed, E-mail: memmm67@yahoo.com [Reactors Department, Nuclear Researches Center, Inshas (Egypt); Shaat, M. [Reactors Department, Nuclear Researches Center, Inshas (Egypt); Kady, M. El [Mechanical Power Engineering Department, Faculty of Engineering, Al Azhar University, Cairo (Egypt)

    2016-04-15

    A computer code was established named ET-RR-1-10 to investigate the thermal hydraulic behavior of the ETRR1 (first Egyptian research reactor) research reactor when its power upgraded to 10 MW using the new fuel plate elements type. The work done include both normal and flow reduction conditions. The code modeled primary loop, secondary lop, and reactor kinetics. All code models used finite difference technique. The code results were tested against the available corresponding experimental data taken from a similar research reactor MITR (Massachusetts Institute of Technology research reactor) for the sake of code validation. The results showed good agreement, and the code can be used for thermal hydraulic calculations.

  2. Concept of a BNCT line with in-pool fission converter at MARIA reactor in Swierk

    Science.gov (United States)

    Pytel, Krzysztof; Andrzejewski, Krzysztof; Golnik, Natalia; Osko, Jakub

    2009-01-01

    BNCT facility in the Institute of Atomic Energy in Otwock-Swierk is under construction at the horizontal channel H2 of the research reactor MARIA. Measurements of the neutron energy spectrum performed at the front of the H2 experimental channel, have shown that flux of epithermal neutrons (above 10 keV) at the BNCT irradiation port was below 109 n cm-2 s-1 i.e. it was too low to be directly used for the BNCT treatment. Therefore, a fission converter will be placed between the reactor core and the periphery of the graphite reflector of MARIA reactor. The uranium converter will be powered by the densely packed EK-10 fuel elements with 10% enrichment. Preliminary calculations have shown that the total neutron flux in the converter will be about 1013 n cm-2 s-1 and flux of epithermal neutrons at the entrance to the filter/moderator of the beam will be about 2·1013 n cm-2 s-1.

  3. Characterization of radioactive contaminants and water treatment trials for the Taiwan Research Reactor's spent fuel pool

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Chun-Ping, E-mail: chunping@iner.gov.tw [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China); Lin, Tzung-Yi; Chiao, Ling-Huan; Chen, Hong-Bin [Institute of Nuclear Energy Research, 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan, ROC (China)

    2012-09-30

    Highlights: Black-Right-Pointing-Pointer Deal with a practical radioactive contamination in Taiwan Research Reactor spent fuel pool water. Black-Right-Pointing-Pointer Identify the properties of radioactive contaminants and performance test for water treatment materials. Black-Right-Pointing-Pointer The radioactive solids were primary attributed by ruptured spent fuels, spent resins, and metal debris. Black-Right-Pointing-Pointer The radioactive ions were major composed by uranium and fission products. Black-Right-Pointing-Pointer Diatomite-based ceramic depth filter can simultaneously removal radioactive solids and ions. - Abstract: There were approximately 926 m{sup 3} of water contaminated by fission products and actinides in the Taiwan Research Reactor's spent fuel pool (TRR SFP). The solid and ionic contaminants were thoroughly characterized using radiochemical analyses, scanning electron microscopy equipped with an energy dispersive spectrometer (SEM-EDS), and inductively coupled plasma optical emission spectrometry (ICP-OES) in this study. The sludge was made up of agglomerates contaminated by spent fuel particles. Suspended solids from spent ion-exchange resins interfered with the clarity of the water. In addition, the ionic radionuclides such as {sup 137}Cs, {sup 90}Sr, U, and {alpha}-emitters, present in the water were measured. Various filters and cation-exchange resins were employed for water treatment trials, and the results indicated that the solid and ionic contaminants could be effectively removed through the use of <0.9 {mu}m filters and cation exchange resins, respectively. Interestingly, the removal of U was obviously efficient by cation exchange resin, and the ceramic depth filter composed of diatomite exhibited the properties of both filtration and adsorption. It was found that the ceramic depth filter could adsorb {beta}-emitters, {alpha}-emitters, and uranium ions. The diatomite-based ceramic depth filter was able to simultaneously

  4. Positron Annihilation Studies of VVER Type Reactor Steels

    OpenAIRE

    Brauer, G.

    1995-01-01

    A summary of recent positron annihilation work on Russian VVER type reactor steels is presented. Thereby, special attention is paid to the outline of basic processes that might help to understand the positron behaviour in this class of industrial material. The idea of positron trapping by irradiation-induced precipitates, which are probably carbides, is discussed in detail.

  5. Simulation Research on Decay Heat Removal System in Primary Loop of Pool-type Sodium-cooled Fast Reactor%池式钠冷快堆事故余热排出系统一回路仿真研究

    Institute of Scientific and Technical Information of China (English)

    姜博; 张智刚; 于洋; 陈广亮; 张志俭

    2015-01-01

    池式钠冷快堆事故余热排出系统采用了非能动工作原理,依靠液态钠及空气的自然对流排出堆芯余热。为研究事故工况下余热排出系统一回路的换热能力,基于 FORTRAN 语言,建立堆芯单通道及盒间流模型,采用全隐二阶迎风差分格式及改进的欧拉法离散求解,对事故余热排出系统一回路系统进行数值模拟,并对全厂断电事故进行仿真计算验证。结果表明:该程序能较好地反映事故余热排出系统瞬态变化过程,并可达到超实时仿真。%T he decay heat removal system in pool‐type sodium‐cooled fast reactor (PSFR) is the passive safety system ,which depends on the natural circulation of sodium and air to keep the reactor coolant cooled .In order to verify the characteristics of the heat transfer of decay heat removal system in primary loop for accident condition ,the core single‐channel model and the flow between fuel assemblies model were established to simulate the decay heat removal system of primary loop and testify the program on station blackout accident , by using fully‐implicit second‐order upwind scheme and ameliorative Eular method to solve the equations based on FORTRAN .The calculation results show that the program could reflect the transient characteristics of the decay heat removal system ,and it could reach excess real‐time simulation .

  6. Deterministic Analysis of a Beyond Design Basis Accident in a Low Power, Pin-Type Fuel Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nagah Abdou, Hesham Mohammed [INVAP S. E., Bariloche (Argentina)

    2013-07-01

    A Beyond Design Basis Accident has been analyzed for a pool type research reactor with pin-type, Zry4 clad fuel. This is a low power research reactor (maximum power: 100kW) with neutron beam facilities. Two scenarios are considered: a neutron beam rapture that results in a fraction of the core submerged in water and a catastrophic failure that results in a fully uncovered core. The paper discusses the different cooling mechanisms for these two BDBAs and compares results for both scenarios, with predictions of no core damage in any situation. Core damage is defined as CHFR↔1.5 and/or Tclad→T start of breakaway oxidation temperature. In addition, the paper compares calculations with a thermalhydraulic code and an analytical model. This paper allows to analyze the applicability of regular thermalhydraulic codes to BDBA accident scenarios in low power research reactors.

  7. Diatomite Type Filters for Swimming Pools. Standard No. 9, Revised October, 1966.

    Science.gov (United States)

    National Sanitation Foundation, Ann Arbor, MI.

    Pressure and vacuum diatomite type filters are covered in this standard. The filters herein described are intended to be designed and used specifically for swimming pool water filtration, both public and residential. Included are the basic components which are a necessary part of the diatomite type filter such as filter housing, element supports,…

  8. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  9. An innovative reactor-type biosensor for BOD rapid measurement.

    Science.gov (United States)

    Wang, Jianlong; Zhang, Yixin; Wang, Yeyao; Xu, Runhua; Sun, Zhonghua; Jie, Zhou

    2010-03-15

    Biochemical oxygen demand (BOD) is one of the most important and widely used parameters for characterizing the organic pollution of water and wastewater. In this paper, a novel reactor-type biosensor for rapid measurement of BOD was developed, based on using immobilized microbial cell (IMC) beads as recognition bio-element in a completely mixed reactor which was used as determining chamber, replacing the traditionally used membrane as recognition bio-element. The IMC beads were freely suspended in the aqueous solution, so the mass transfer resistance for dissolved oxygen and organic compounds significantly reduced, and the quantity of the microbial cells used as recognition element can be easily adjusted, in comparison with the traditional membrane-type BOD biosensor, in which exists a unadjustable contradiction between the quantity of biomass and the thickness of the bio-membrane, thus limiting the stability and the detection limit. This novel kind of BOD biosensor significantly increased the sensitivity of the response, the detecting precision and prolonged the life time of the recognition element. The experimental data showed that the most appropriate temperature for biochemical reaction in the reactor was 30 degrees C, and the IMC beads could keep the bioactivity for about 70d at the detecting frequency of 8 times every day. The standard deviation of repeatability and the reproducibility of responses were within +/-6.4% and +/-5.0%, respectively, which are within acceptable bias limits, and meet the requirement of BOD rapid measurement.

  10. How to project onto the monotone nonnegative cone using Pool Adjacent Violators type algorithms

    CERN Document Server

    Németh, A B

    2012-01-01

    The metric projection onto an order nonnegative cone from the metric projection onto the corresponding order cone is derived. Particularly, we can use Pool Adjacent Violators-type algorithms developed for projecting onto the monotone cone for projecting onto the monotone nonnegative cone too.

  11. Birth order and childhood type 1 diabetes risk: a pooled analysis of 31 observational studies

    DEFF Research Database (Denmark)

    Cardwell, Chris R; Stene, Lars C; Joner, Geir

    2011-01-01

    The incidence rates of childhood onset type 1 diabetes are almost universally increasing across the globe but the aetiology of the disease remains largely unknown. We investigated whether birth order is associated with the risk of childhood diabetes by performing a pooled analysis of previous...

  12. Selection of Type I and Type II Methanotrophic Proteobacteria in a Fluidized Bed Reactor under Non-Sterile Conditions

    Science.gov (United States)

    2011-08-01

    00-00-2011 to 00-00-2011 4. TITLE AND SUBTITLE Selection of Type I and Type II methanotrophic proteobacteria in a fluidized bed reactor under...laboratory- scale fluidized bed reactor was initially inoculated with a Type II Methylocystis-like dominated culture. At elevated levels of dissolved...personal copy Selection of Type I and Type II methanotrophic proteobacteria in a fluidized bed reactor under non-sterile conditions Andrew R. Pfluger a, Wei

  13. Description of the magnox type of gas cooled reactor (MAGNOX)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

    The present report comprises a technical description of the MAGNOX type of reactor as it has been build in Great Britain. The Magnox reactor is gas cooled (CO{sub 2}) with graphite moderators. The fuels is natural uranium in metallic form, canned with a magnesium alloy called 'Magnox'. The Calder Hall Magnox plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other stations are given in tables with a summary of design data. Special design features are also shortly described. Where specific data for Calder Hall Magnox has not been available, corresponding data from other Magnox plants has been used. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 sub-project 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au)

  14. Development of toroid-type HTS DC reactor series for HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon; Yu, In-Keun [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2015-11-15

    Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  15. Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

    Energy Technology Data Exchange (ETDEWEB)

    Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

    2005-07-01

    In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

  16. Scale Effects on Magnet Systems of Heliotron-Type Reactors

    Institute of Scientific and Technical Information of China (English)

    S. Imagawa; A. Sagara

    2005-01-01

    For power plants heliotron-type reactors have attractive advantages, such as no current-disruptions, no current-drive, and wide space between helical coils for the maintenance of in-vessel components. However, one disadvantage is that a major radius has to be large enough to obtain large Q-value or to produce sufficient space for blankets. Although the larger radius is considered to increase the construction cost, the influence has not been understood clearly,yet. Scale effects on superconducting magnet systems have been estimated under the conditions of a constant energy confinement time and similar geometrical parameters. Since the necessary magnetic field with a larger radius becomes lower, the increase rate of the weight of the coil support to the major radius is less than the square root. The necessary major radius will be determined mainly by the blanket space. The appropriate major radius will be around 13 m for a reactor similar to the Large Helical Device (LHD).

  17. Development of toroid-type HTS DC reactor series for HVDC system

    Science.gov (United States)

    Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

    2015-11-01

    This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

  18. LMFBR type reactor and power generation system using the same

    Energy Technology Data Exchange (ETDEWEB)

    Otsubo, Akira.

    1994-02-25

    A reactor core void reactivity of a reactor main body is set to negative or zero. A heat insulation structure is disposed on the inner wall surface of a reactor container. Oxide fuels or nitride fuels are used. A fuel pin cladding tube has a double walled structure having an outer side of stainless steel and an inner side of niobium alloy. Upon imaginary event, boiling is allowed. Even if boiling of coolants should occur by temperature elevation of fuels upon imaginary event, since reactor core fuels comprises oxides or nitrides, they have a heat resistance, further, and since the fuel pin cladding tube has super heat resistance, it has a high temperature strength, so that it is not ruptured and durable to the coolant boiling temperature. Since the reactor core void reactivity is negative or zero, the reactor core is in a subcritical state by the boiling, and the reactor core power is reduced to several % of the rated power. Accordingly, boiling and non-boiling are repeated substantially permanently in the reactor core, during which safety can be kept with no operator's handling. Further, heat generated in the reactor core is gradually removed by an air cooling system for the reactor container. (N.H.).

  19. Biodiversity of 52 chicken populations assessed by microsatellite typing of DNA pools

    Directory of Open Access Journals (Sweden)

    Thomson Pippa

    2003-09-01

    Full Text Available Abstract In a project on the biodiversity of chickens funded by the European Commission (EC, eight laboratories collaborated to assess the genetic variation within and between 52 populations from a wide range of chicken types. Twenty-two di-nucleotide microsatellite markers were used to genotype DNA pools of 50 birds from each population. The polymorphism measures for the average, the least polymorphic population (inbred C line and the most polymorphic population (Gallus gallus spadiceus were, respectively, as follows: number of alleles per locus, per population: 3.5, 1.3 and 5.2; average gene diversity across markers: 0.47, 0.05 and 0.64; and proportion of polymorphic markers: 0.91, 0.25 and 1.0. These were in good agreement with the breeding history of the populations. For instance, unselected populations were found to be more polymorphic than selected breeds such as layers. Thus DNA pools are effective in the preliminary assessment of genetic variation of populations and markers. Mean genetic distance indicates the extent to which a given population shares its genetic diversity with that of the whole tested gene pool and is a useful criterion for conservation of diversity. The distribution of population-specific (private alleles and the amount of genetic variation shared among populations supports the hypothesis that the red jungle fowl is the main progenitor of the domesticated chicken.

  20. Hydrogeology, ground-water quality, and the possible effects of a hypothetical radioactive water spill, Plainsboro Township, New Jersey

    Science.gov (United States)

    Lewis, J.C.; Spitz, F.J.

    1987-01-01

    Princeton University, under contract to the Department of Energy , maintains a Tokamak fusion test reactor in New Jersey. The U.S. Geological Survey investigated groundwater flow and estimated the effects of a hypothetical spill of radioactive water at the site on the local groundwater system. The study included test drilling; aquifer testing; measurement of water levels, infiltration capacity, and stream discharge; and a simulation of the hypothetical spill. The Triassic Stockton Formation-a water supply aquifer composed primarily of jointed siltstone and sandstone-underlies the site. The aquifer is confined by overlying weathered bedrock and underlying unjointed rock. Weathered bedrock is overlain by unconsolidated, partially saturated material which ranges from 6 to 39 ft in thickness. Groundwater recharge is by lateral flow into the study area, stream leakage, and precipitation. Discharge is by pumpage, evapotranspiration, stream inflow, and lateral flow out of the study area. Transmissivity of the aquifer is about 1,740 sq ft/day, and the storage coefficient is about 0.0002. The average linear velocity of groundwater at the site ranges from 100 to 270 ft/yr depending on location and time of year. The velocity over a large part of the site is controlled by on-site pumpage. Groundwater samples were collected and analyzed for common ions, trace metals, and tritium. The analyses reported no concentrations of common ions or trace metals which exceeded the criteria for drinking water standards recommended by the EPA, except for some instances of moderately high concentrations of iron and manganese. Iron and manganese are common in groundwater and surface water in the area and are not indicative of an on-site source of contamination. Tritium concentrations in the collected samples were also considered representative of background levels and were well below the maximum concentration permitted by the EPA. The fate of spilled radioactive water after a hypothetical

  1. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950`s are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  2. Reactor Materials Program: Mechanical properties of irradiated Types 304 and 304L stainless steel weldment components

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.; Caskey, G.R. Jr.

    1991-12-01

    The vessels (reactor tanks) of the Savannah River Site nuclear production reactors constructed in the 1950's are comprised of Type 304 stainless steel with Type 308 stainless steel weld filler. Irradiation exposure to the reactor tank sidewalls through reactor operation has caused a change in the mechanical properties of these materials. A database of as-irradiated mechanical properties for site-specific materials and irradiation conditions has been produced for reactor tank structural analyses and to quantify the effects of radiation-induced materials degradation for evaluating reactor service life. The data has been collected from the SRL Reactor Materials Program (RMP) irradiations and testing of archival stainless steel weldment components and from previous SRL programs to measure properties of irradiated reactor Thermal Shield weldments and reactor tank (R-tank) sidewall material. Irradiation programs of the RMP are designed to quantify mechanical properties at tank operating temperatures following irradiation to present and future tank wall maximum exposure conditions. The exposure conditions are characterized in terms of fast neutron fluence (E{sub n} > 0.1 MeV) and displacements per atom (dpa){sup 3}. Tensile properties, Charpy-V notch toughness, and elastic-plastic fracture toughness were measured for base, weld, and weld heat-affected zone (HAZ) weldment components from archival piping specimens following a Screening Irradiation in the University of Buffalo Reactor (UBR) and following a Full-Term Irradiation in the High Flux Isotope Reactor (HFIR).

  3. Investigation of the thermal performance of a vertical two-phase closed thermosyphon as a passive cooling system for a nuclear reactor spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Kusuma, Mukhsinun Hadi; Putra, Nandy; Imawan, Ficky Augusta [Heat Transfer Laboratory, Department of Mechanical Engineering Universitas Indonesia, Kampus (Indonesia); Antariksawan, Anhar Riza [Centre for Nuclear Reactor Safety and Technology, National Nuclear Energy Agency of Indonesia (BATAN), Kawasan Puspiptek Serpong (Indonesia)

    2017-04-15

    The decay heat that is produced by nuclear reactor spent fuel must be cooled in a spent fuel storage pool. A wickless heat pipe or a vertical two-phase closed thermosyphon (TPCT) is used to remove this decay heat. The objective of this research is to investigate the thermal performance of a prototype model for a large-scale vertical TPCT as a passive cooling system for a nuclear research reactor spent fuel storage pool. An experimental investigation and numerical simulation using RELAP5/MOD 3.2 were used to investigate the TPCT thermal performance. The effects of the initial pressure, filling ratio, and heat load were analyzed. Demineralized water was used as the TPCT working fluid. The cooled water was circulated in the water jacket as a cooling system. The experimental results show that the best thermal performance was obtained at a thermal resistance of 0.22°C/W, the lowest initial pressure, a filling ratio of 60%, and a high evaporator heat load. The simulation model that was experimentally validated showed a pattern and trend line similar to those of the experiment and can be used to predict the heat transfer phenomena of TPCT with varying inputs.

  4. Birth order and childhood type 1 diabetes risk: a pooled analysis of 31 observational studies

    DEFF Research Database (Denmark)

    Cardwell, Chris R; Stene, Lars C; Joner, Geir

    2010-01-01

    BACKGROUND: The incidence rates of childhood onset type 1 diabetes are almost universally increasing across the globe but the aetiology of the disease remains largely unknown. We investigated whether birth order is associated with the risk of childhood diabetes by performing a pooled analysis...... and after adjustment for confounders, and investigate heterogeneity. RESULTS: Data were available for 6 cohort and 25 case-control studies, including 11¿955 cases of type 1 diabetes. Overall, there was no evidence of an association prior to adjustment for confounders. After adjustment for maternal age...... at birth and other confounders, a reduction in the risk of diabetes in second- or later born children became apparent [fully adjusted OR¿=¿0.90 95% confidence interval (CI) 0.83-0.98; P¿=¿0.02] but this association varied markedly between studies (I(2)¿=¿67%). An a priori subgroup analysis showed...

  5. IRL murrab jõuliselt Rahvaliitu kuuluvaid omavalitsusjuhte / Kärt Anvelt

    Index Scriptorium Estoniae

    Anvelt, Kärt, 1973-

    2011-01-01

    Isamaa ja Res Publica Liit eesotsas ministrite Juhan Partsi, Siim-Valmar Kiisleri ja detsembri alguses erakonnaga ühinenud Aivar Kokaga on asunud meelitama enda poole üle tulema Rahvaliitu kuuluvaid vallavanemaid. Regionaalminister Siim-Valmar Kiisleri kommentaar

  6. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  7. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  8. Reactor pressure vessels as type B transport containment boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Nickell, R.E. [Applied Science and Technology, Inc., Poway, CA (United States); Griesbach, T.J. [ATI Consulting, Danville, CA (United States)

    1998-07-01

    Transportation risk and personnel exposure, as well as the cost of decommissioning nuclear power plants, can all be reduced significantly through the one-time use of the reactor pressure vessel as a containment boundary for shipping the activated internal components from the reactor site to a burial site. In order to help provide the technical basis for this end-use application, the ASME Board on Nuclear Codes and Standards, through its Subcommittee XI, has prepared a draft nuclear code case that contains requirements for any modifications to the vessel, including materials, design, fabrication, and examination. In particular, the requirements for evaluation of potential brittle fracture as the result of potentially low ambient shipping temperatures combined with hypothetical transportation accident loading are addressed. Existing ASME Code Section XI rules for linear elastic fracture mechanics evaluation of irradiated reactor pressure vessels have been adapted and included in the code case. (authors)

  9. Safety and tolerability of sitagliptin in patients with type 2 diabetes: a pooled analysis

    Science.gov (United States)

    Williams-Herman, Debora; Round, Elizabeth; Swern, Arlene S; Musser, Bret; Davies, Michael J; Stein, Peter P; Kaufman, Keith D; Amatruda, John M

    2008-01-01

    Background Sitagliptin, a highly selective dipeptidyl peptidase-4 inhibitor, is the first in a new class of oral antihyperglycemic agents (AHAs) for the treatment of patients with type 2 diabetes. Type 2 diabetes is a life-long disease requiring chronic treatment and management. Therefore, robust assessment of the long-term safety and tolerability of newer therapeutic agents is of importance. The purpose of this analysis was to assess the safety and tolerability of sitagliptin by pooling 12 large, double-blind, Phase IIb and III studies up to 2 years in duration. Methods: This analysis included 6139 patients with type 2 diabetes receiving either sitagliptin 100 mg/day (N = 3415) or a comparator agent (placebo or an active comparator) (N = 2724; non-exposed group). The 12 studies from which this pooled population was drawn represent the double-blind, randomized, Phase IIB and III studies that included patients treated with the clinical dose of sitagliptin (100 mg/day) for at least 18 weeks up to 2 years and that were available in a single safety database as of November 2007. These 12 studies assessed sitagliptin as monotherapy, initial combination therapy with metformin, or add-on combination therapy with other oral AHAs (metformin, pioglitazone, sulfonylurea, sulfonylurea + metformin, or metformin + rosiglitazone). Patients in the non-exposed group were taking placebo, pioglitazone, metformin, sulfonylurea, sulfonylurea + metformin, or metformin + rosiglitazone. This safety analysis used patient-level data from each study to evaluate clinical and laboratory adverse experiences. Results For clinical adverse experiences, the incidence rates of adverse experiences overall, serious adverse experiences, and discontinuations due to adverse experiences were similar in the sitagliptin and non-exposed groups. The incidence rates of specific adverse experiences were also generally similar in the two groups, with the exception of an increased incidence rate of hypoglycemia

  10. Safety and tolerability of sitagliptin in patients with type 2 diabetes: a pooled analysis

    Directory of Open Access Journals (Sweden)

    Davies Michael J

    2008-10-01

    Full Text Available Abstract Background Sitagliptin, a highly selective dipeptidyl peptidase-4 inhibitor, is the first in a new class of oral antihyperglycemic agents (AHAs for the treatment of patients with type 2 diabetes. Type 2 diabetes is a life-long disease requiring chronic treatment and management. Therefore, robust assessment of the long-term safety and tolerability of newer therapeutic agents is of importance. The purpose of this analysis was to assess the safety and tolerability of sitagliptin by pooling 12 large, double-blind, Phase IIb and III studies up to 2 years in duration. Methods: This analysis included 6139 patients with type 2 diabetes receiving either sitagliptin 100 mg/day (N = 3415 or a comparator agent (placebo or an active comparator (N = 2724; non-exposed group. The 12 studies from which this pooled population was drawn represent the double-blind, randomized, Phase IIB and III studies that included patients treated with the clinical dose of sitagliptin (100 mg/day for at least 18 weeks up to 2 years and that were available in a single safety database as of November 2007. These 12 studies assessed sitagliptin as monotherapy, initial combination therapy with metformin, or add-on combination therapy with other oral AHAs (metformin, pioglitazone, sulfonylurea, sulfonylurea + metformin, or metformin + rosiglitazone. Patients in the non-exposed group were taking placebo, pioglitazone, metformin, sulfonylurea, sulfonylurea + metformin, or metformin + rosiglitazone. This safety analysis used patient-level data from each study to evaluate clinical and laboratory adverse experiences. Results For clinical adverse experiences, the incidence rates of adverse experiences overall, serious adverse experiences, and discontinuations due to adverse experiences were similar in the sitagliptin and non-exposed groups. The incidence rates of specific adverse experiences were also generally similar in the two groups, with the exception of an increased incidence

  11. EFFECT OF PARTICLE TYPE ON CYCLONE FORMATION INSIDE A SOLAR REACTOR

    Directory of Open Access Journals (Sweden)

    Min-Hsiu Chien

    2016-07-01

    Full Text Available Solar reactors featuring a circulating cyclone flow pattern provide enhanced heat transfer and longer residence time increasing conversion efficiency. Cyclone flow also works in reducing particle deposition on solar reactor walls and exit which is particularly important issue in solar cracking reactors to avoid clogging. This paper focuses on the physics of cyclone formation inside a solar cracking reactor and experimentally analyzes the effect of particle entrainment on the flow pattern via two dimensional Particle Image Velocimetry (PIV. The cyclone flow structure in the reactor is reconstructed by capturing images from orientations perpendicular or parallel to the geometrical axis of the reactor. In order to conduct PIV measurements and to reconstruct the cyclone structure inside the solar reactor, the experiment was operated at room temperature with the flow configuration matching that of a solar reactor operating at high temperatures. Two types of seeding particles were tested, namely tri-ethylene glycol (TEG and solid carbon. The effectiveness of the screening flow was evaluated by measuring the quantity of solid particles deposit on the reactor walls. The Stokes flow analysis of each particle species was performed and the cyclone vector fields generated by using different particles are compared.

  12. Variation of biomass and carbon pools with forest type in temperate forests of Kashmir Himalaya, India.

    Science.gov (United States)

    Dar, Javid Ahmad; Sundarapandian, Somaiah

    2015-02-01

    An accurate characterization of tree, understory, deadwood, floor litter, and soil organic carbon (SOC) pools in temperate forest ecosystems is important to estimate their contribution to global carbon (C) stocks. However, this information on temperate forests of the Himalayas is lacking and fragmented. In this study, we measured C stocks of tree (aboveground and belowground biomass), understory (shrubs and herbaceous), deadwood (standing and fallen trees and stumps), floor litter, and soil from 111 plots of 50 m × 50 m each, in seven forest types: Populus deltoides (PD), Juglans regia (JR), Cedrus deodara (CD), Pinus wallichiana (PW), mixed coniferous (MC), Abies pindrow (AP), and Betula utilis (BU) in temperate forests of Kashmir Himalaya, India. The main objective of the present study is to quantify the ecosystem C pool in these seven forest types. The results showed that the tree biomass ranged from 100.8 Mg ha(-1) in BU forest to 294.8 Mg ha(-1) for the AP forest. The understory biomass ranged from 0.16 Mg ha(-1) in PD forest to 2.36 Mg ha(-1) in PW forest. Deadwood biomass ranged from 1.5 Mg ha(-1) in PD forest to 14.9 Mg ha(-1) for the AP forest, whereas forest floor litter ranged from 2.5 Mg ha(-1) in BU and JR forests to 3.1 Mg ha(-1) in MC forest. The total ecosystem carbon stocks varied from 112.5 to 205.7 Mg C ha(-1) across all the forest types. The C stocks of tree, understory, deadwood, litter, and soil ranged from 45.4 to 135.6, 0.08 to 1.18, 0.7 to 6.8, 1.1 to 1.4, and 39.1-91.4 Mg ha(-1), respectively, which accounted for 61.3, 0.2, 1.4, 0.8, and 36.3 % of the total carbon stock. BU forest accounted 65 % from soil C and 35 % from biomass, whereas PD forest contributed only 26 % from soil C and 74 % from biomass. Of the total C stock in the 0-30-cm soil, about 55 % was stored in the upper 0-10 cm. Soil C stocks in BU forest were significantly higher than those in other forests. The variability of C pools of different ecosystem components is

  13. Study on Thermal-Hydraulic Behavior of an Integral Type Reactor under Heaving Condition

    OpenAIRE

    2014-01-01

    A self-developed program was used to study the thermal-hydraulic behavior of an integral type reactor under heaving condition. Comparison of calculated results with the data of experiments performed on a natural circulation loop designed with reference to an integral type reactor of Tsinghua University in inclination, heaving, and rolling motions was carried out. Characteristics of natural circulation in heaving motion and effect of motion parameters on natural circulation were investigated. ...

  14. The spatial variability of water chemistry and DOC in bog pools: the importance of slope position, diurnal turnover and pool type

    Science.gov (United States)

    Holden, Joseph; Turner, Ed; Baird, Andy; Beadle, Jeannie; Billett, Mike; Brown, Lee; Chapman, Pippa; Dinsmore, Kerry; Dooling, Gemma; Grayson, Richard; Moody, Catherine; Gee, Clare

    2017-04-01

    We have previously shown that marine influence is an important factor controlling regional variability of pool water chemistry in blanket peatlands. Here we examine within-site controls on pool water chemistry. We surveyed natural and artificial (restoration sites) bog pools at blanket peatland sites in northern Scotland and Sweden. DOC, pH, conductivity, dissolved oxygen, temperature, cations, anions and absorbance spectra from 220-750nm were sampled. We sampled changes over time but also conducted intensive spatial surveys within individual pools and between pools on the same sampling days at individual study sites. Artificial pools had significantly greater DOC concentrations and different spectral absorbance characteristics when compared to natural pools at all sites studied. Within-pool variability in water chemistry tended to be small, even for very large pools ( 400 m2), except where pools had a layer of loose, mobile detritus on their beds. In these instances rapid changes took place between the overlying water column and the mobile sediment layer wherein dissolved oxygen concentrations dropped from values of around 12-10 mg/L to values less than 0.5 mg/L over just 2-3 cm of the depth profile. Such strong contrasts were not observed for pools which had a hard peat floor and which lacked a significant detritus layer. Strong diurnal turnover occurred within the pools on summer days, including within small, shallow pools (e.g. dissolved oxygen concentrations which originated at the surface and was then cycled downwards as the pool surface waters cooled. Slope location was a significant control on several pool water chemistry variables including pH and DOC concentration with accumulation (higher concentrations) in pools that were located further downslope in both natural and artificial pool systems. These processes have important implications for our interpretation of water chemistry and gas flux data from pool systems, how we design our sampling strategies and

  15. Sensitivity Analysis on LOCCW of Westinghouse typed Reactors Considering WOG2000 RCP Seal Leakage Model

    Energy Technology Data Exchange (ETDEWEB)

    Na, Jang-Hwan; Jeon, Ho-Jun; Hwang, Seok-Won [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, we focus on risk insights of Westinghouse typed reactors. We identified that Reactor Coolant Pump (RCP) seal integrity is the most important contributor to Core Damage Frequency (CDF). As we reflected the latest technical report; WCAP-15603(Rev. 1-A), 'WOG2000 RCP Seal Leakage Model for Westinghouse PWRs' instead of the old version, RCP seal integrity became more important to Westinghouse typed reactors. After Fukushima accidents, Korea Hydro and Nuclear Power (KHNP) decided to develop Low Power and Shutdown (LPSD) Probabilistic Safety Assessment (PSA) models and upgrade full power PSA models of all operating Nuclear Power Plants (NPPs). As for upgrading full power PSA models, we have tried to standardize the methodology of CCF (Common Cause Failure) and HRA (Human Reliability Analysis), which are the most influential factors to risk measures of NPPs. Also, we have reviewed and reflected the latest operating experiences, reliability data sources and technical methods to improve the quality of PSA models. KHNP has operating various types of reactors; Optimized Pressurized Reactor (OPR) 1000, CANDU, Framatome and Westinghouse. So, one of the most challengeable missions is to keep the balance of risk contributors of all types of reactors. This paper presents the method of new RCP seal leakage model and the sensitivity analysis results from applying the detailed method to PSA models of Westinghouse typed reference reactors. To perform the sensitivity analysis on LOCCW of the reference Westinghouse typed reactors, we reviewed WOG2000 RCP seal leakage model and developed the detailed event tree of LOCCW considering all scenarios of RCP seal failures. Also, we performed HRA based on the T/H analysis by using the leakage rates for each scenario. We could recognize that HRA was the sensitive contributor to CDF, and the RCP seal failure scenario of 182gpm leakage rate was estimated as the most important scenario.

  16. Concept of magnet systems for LHD-type reactor

    Science.gov (United States)

    Imagawa, S.; Takahata, K.; Tamura, H.; Yanagi, N.; Mito, T.; Obana, T.; Sagara, A.

    2009-07-01

    Heliotron reactors have attractive features for fusion power plants such as having no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered to be the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, a major radius of plasma of 14-17 m with a central toroidal field of 6-4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120-140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress are comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than the 1000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is more than 150 m, that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with a small extension of the ITER technology.

  17. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Chi Thanh

    2009-09-15

    Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents. In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment. The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis. The CFD method, on the one hand

  18. Pool-type fishways: two different morpho-ecological cyprinid species facing plunging and streaming flows.

    Science.gov (United States)

    Branco, Paulo; Santos, José M; Katopodis, Christos; Pinheiro, António; Ferreira, Maria T

    2013-01-01

    Fish are particularly sensitive to connectivity loss as their ability to reach spawning grounds is seriously affected. The most common way to circumvent a barrier to longitudinal connectivity, and to mitigate its impacts, is to implement a fish passage device. However, these structures are often non-effective for species with different morphological and ecological characteristics so there is a need to determine optimum dimensioning values and hydraulic parameters. The aim of this work is to study the behaviour and performance of two species with different ecological characteristics (Iberian barbel Luciobarbus bocagei-bottom oriented, and Iberian chub Squalius pyrenaicus-water column) in a full-scale experimental pool-type fishway that offers two different flow regimes-plunging and streaming. Results showed that both species passed through the surface notch more readily during streaming flow than during plunging flow. The surface oriented species used the surface notch more readily in streaming flow, and both species were more successful in moving upstream in streaming flow than in plunging flow. Streaming flow enhances upstream movement of both species, and seems the most suitable for fishways in river systems where a wide range of fish morpho-ecological traits are found.

  19. Dependence of the characteristics of bubbles on types of sonochemical reactors.

    Science.gov (United States)

    Yasui, Kyuichi; Tuziuti, Toru; Iida, Yasuo

    2005-01-01

    Computer simulations of bubble oscillations in liquid water irradiated by an ultrasonic wave have revealed that the characteristic of bubbles depends on types of sonochemical reactors: a horn-type reactor and a standing-wave type reactor. When the acoustic amplitude is large at 20 kHz, the bubble content is mostly water vapor even at the end of the bubble collapse and the temperature inside a bubble at the collapse is relatively low. On the other hand, when the acoustic amplitude is relatively low, the bubble content is mostly noncondensable gas at the end of the bubble collapse and the bubble temperature is relatively high. In a horn-type sonochemical reactor, the former type of bubbles are dominant because many bubbles exist near the horn-tip where the acoustic amplitude is large, while in a standing-wave type reactor the latter type of bubbles are dominant because the Bjerknes force gathers bubbles at a region where acoustic amplitude is relatively low.

  20. Dynamic Response of VVER 1000 Type Reactor Excited by Pressure Pulsations

    OpenAIRE

    Zeman, Vladimír; Hlaváč, Zdeněk

    2008-01-01

    The paper deals with the modelling of forced vibrations of reactor components excited by pressure pulsations generated by main circulation pumps. For the vibration analysis a new generalised model of the reactor with spatial localization of the nuclear fuel assemblies and protection tubes, continuously mass distribution of beam type components and more accurate model of the linear stepper drives for actuation of control cassettes was applied. Slightly different pump revolutions are sources of...

  1. Analysis of the MEX-15 multipurpose reactor using SRAC code system

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1992-12-15

    The MEX-15 is a conceptual design of a Multipurpose Reactor with thermal power of 15 MW and this reactor is pool type with fuel plates U{sub 3}0{sub 8}-Al of low enrichment uranium. This report presents the static calculation for the MEX-15 reactor using SRAC code system and was developed under the collaboration agreement between ININ-JAERI in Research Reactor Technology Development Division of Department of Research Reactor in Tokai Research Establishment. (Author)

  2. Modification of Neutron Kinetic Code for Plate Type Fuel Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Salah Ud-Din Khan

    2013-01-01

    Full Text Available The research is conducted on the modification of neutron kinetic code for the plate type fuel nuclear reactor. REMARK is a neutron kinetic code that works only for the cylindrical type fuel nuclear reactor. In this research, our main emphasis is on the modification of this code in order to be applicable for the plate type fuel nuclear reactor. For this purpose, detailed mathematical studies have been performed and are subjected to write the program in Fortran language. Since REMARK code is written in Fortran language, so we have developed the program in Fortran and then inserted it into the source library of the code. The main emphasis is on the modification of subroutine in the source library of the code for hexagonal fuel assemblies with plate type fuel elements in it. The number of steps involved in the modification of the code has been included in the paper. The verification studies were performed by considering the small modular reactor with hexagonal assemblies and plate type fuel in it to find out the power distribution of the reactor core. The purpose of the research is to make the code work for the hexagonal fuel assemblies with plate type fuel element.

  3. LMFBR type reactor core and its fuel exchange method

    Energy Technology Data Exchange (ETDEWEB)

    Ishibashi, Yoko; Koyama, Jun-ichi; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro

    1996-08-20

    Upon initial loading, two kinds of fuel assemblies including first fuel assemblies having a highest enrichment degree and second fuel assemblies having a lowest enrichment degree are loaded. The average fuel enrichment degree of an upper region of the first fuel assembly is made greater than that of the lower region. The reactivity of the lower region of the first fuel assembly is made lower than that of the upper portion to reduce power peak. Upon transfer from a first cycle to a second cycle, at least one of the second fuel assemblies is exchanged by the same number of the third fuel assemblies. In this case, an average fuel enrichment degree of the upper region of the third fuel assembly is made greater than that of the lower region to suppress the reactivity in the lower region of the third fuel assembly lower than the reactivity in the upper region thereby reducing the power peak. Thus, the upper power peak over the entire reactor core is moderated thereby capable of ensuring the reactor shut down margin without deteriorating the same. (N.H.)

  4. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  5. Analysis of simulation results of damaged nuclear fuel accidents at NPPs with shell-type nuclear reactors

    Directory of Open Access Journals (Sweden)

    Igor L. Kozlov

    2015-03-01

    Full Text Available Lessons from the accident at the Fukushima Daiichi NPP made it necessary to reevaluate and intensificate the work on modeling and analyzing various scenarios of severe accidents with damage to the nuclear fuel in the reactor, containment and spent nuclear fuel storage pool with the expansion of the primary initiating event causes group listing. Further development of computational tools for modeling the explosion prevention criteria as to steam and gas mixtures, considering the specific thermal-hydrodynamic conditions and mechanisms of explosive situations arrival at different stages of a severe accident development, is substantiated. Based on the analysis of the known shell-type nuclear reactors accidents results the explosion safety thermodynamic criteria are presented, the parameters defining the steam and gas explosions conditions are found, the need to perform the further verification and validation of deterministic codes serving to simulate general accident processes behavior as well as phase-to-phase interaction calculated dependencies is established. The main parameters controlling and defining the criteria explosion safety effective regulation areas and their optimization conditions are found.

  6. Fast Traveling-Wave Reactor of the Channel Type

    CERN Document Server

    Rusov, Vitaliy D; Vashchenko, Volodymyr N; Chernezhenko, Sergei A; Kakaev, Andrei A; Pantak, Oksana I

    2015-01-01

    The main aim of this paper is to solve the technological problems of the TWR based on the technical concept described in our priority of invention reference, which makes it impossible, in particular, for the fuel claddings damaging doses of fast neutrons to excess the ~200 dpa limit. Thus the essence of the technical concept is to provide a given neutron flux at the fuel claddings by setting the appropriate speed of the fuel motion relative to the nuclear burning wave. The basic design of the fast uranium-plutonium nuclear traveling-wave reactor with a softened neutron spectrum is developed, which solves the problem of the radiation resistance of the fuel claddings material.

  7. Study on the selection of nuclear fuel type for a hybrid power extraction reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Dong Han; Park, Won Suk [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-11-01

    The development of a subcritical transmutation reactor concept is emerging for reducing the amounts of actinides and long-lived nuclides in the spent fuel from nuclear power plants. This technology may make contribution to reduce the human risks associated with constructing radio-waste disposal facilities. One of the important issues for the design of the reactor is the selection of a suitable nuclear fuel type. Choosing the best nuclear fuel type for the reactor may not be easy since there exist several criteria associated with neutronic aspects, thermal performance, safety problem, cost problem, radiation damage in the reactor, etc. The best option should be chosen based on the maximization of our needs in this situation. This study presents a logical decision model for this issue using an analytic hierarchy process (AHP). Hierarchy is a representation of a system to study the functional relations of its components and its impact on the entire system. The study shows first how to construct hierarchy representing their relations and then measure the individual element's impact to the entire system for a quantitative decision making. Current four fuel types; metal, oxide, molten salt, and nitride, were selected and analyzed based on several characteristics with respect to overall comparison. Based on the decision model developed, the study concludes that the metal fuel type is the best choice for the transmutation reactor. The proposed approach is intended to help people be rational and logical in making decisions such complex task. 13 refs., 16 figs., 16 tabs. (Author)

  8. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K. [Changwon National University, Changwon (Korea, Republic of); Kim, J. H. [Daejeon University, Daejeon (Korea, Republic of)

    2015-03-15

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed.

  9. Study on the selection of nuclear fuel type for a hybrid power extration reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, D. H.; Park, W. S. [KAERI, Taejon (Korea, Republic of)

    1999-05-01

    In order to solve the problem related to long-lived radioactive nuclides in spent fuel, development of a subcritical transmutation reactor concept is emerging. One of the important issues for the design of the reactor may be the selection of a suitable nuclear fuel type. This study presents a logical decision model for this issue using an analytic hierachy process (AHP). Hierarchy is a representation of a system to study the functional relations of its components and its impact on the entire system. The study shows first how to construct hierachy representing their relations and then measure the individual element's impact to the entire system for a quantitative decision making. Current four fuel types; metal, oxide, molten salt, and nitride, were selected and analyzed based on several characteristics with respect to overall comparison. Based on the decision model, the study concludes that the metal fuel type is the best choice for the transmutation reactor.

  10. Development of a Monolithic Research Reactor Fuel Type at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Clark, C.R.; Briggs, R.J.

    2004-10-06

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been tasked with the conversion of research reactors from highly enriched to low-enriched uranium (LEU). To convert several high power reactors, monolithic fuel, a new fuel type, is being developed. This fuel type replaces the standard fuel dispersion with a fuel alloy foil, which allows for fuel densities far in excess of that found in dispersion fuel. The single-piece fuel foil also contains a significantly lower interface area between the fuel and the aluminum in the plate than the standard fuel type, limiting the amount of detrimental fuel-aluminum interaction that can occur. Implementation of monolithic fuel is dependant on the development of a suitable fabrication method as traditional roll-bonding techniques are inadequate.

  11. Use of Stable Noble Gases as a Predictor of Reactor Fuel Type and Exposure

    Energy Technology Data Exchange (ETDEWEB)

    Fearey, B.L.; Charlton, W.S.; Perry, R.T.; Poths, J.; Wilson, W.B.; Hemberger, P.H.; Nakhleh, C.W.; Stanbro, W.D.

    1999-08-30

    Ensuring spent reactor fuel is not produced to provide weapons-grade plutonium is becoming a major concern as many countries resort to nuclear power as a solution to their energy problems. Proposed solutions range from the development of proliferation resistant fuel to continuous monitoring of the fuel. This paper discusses the use of the stable isotopes of the fissiogenic noble gases, xenon and krypton, for determining the burnup characteristics, fuel type, and the reactor type of the fuel from which the sample was obtained. The gases would be collected on-stack as the fuel is reprocessed, and thus confirm that the fuel is as declared.

  12. A comparative study of the attenuation of reactor thermal neutrons in different types of concrete

    Energy Technology Data Exchange (ETDEWEB)

    Bashiter, I.I. [Zagazig Univ. (Egypt). Dept. of Physics; El-Sayed Abdo, A.; Makarious, A.S. [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Centre

    1996-05-20

    This study was carried out to assess the distribution of thermal neutrons emitted directly from the core of the ET-RR-1 reactor in ordinary concrete, ilmenite concrete and ilmenite-limonite concrete shields. Measurements were carried out by using a direct beam and a cadmium filtered beam of reactor neutrons. The neutron dose distributions were measured using Li{sub 2}B{sub 4}O{sub 7}:Mn thermoluminescent dosimeters. The data obtained show that ilmenite concrete is better for slow and thermal neutron attenuation than both ordinary and ilmenite-limonite concrete. Also it was concluded that thermal neutrons emitted directly from the reactor core are highly absorbed within the first few centimeters of each type of concrete. The thickness of ilmenite concrete required to attenuate the doses of neutrons to a certain value along the beam axis for a direct reactor beam estimated to be about 75 and 57% of the shield thickness made from ordinary and ilmenite-limonite concretes, respectively. Empirical formulae were derived to calculate the neutron dose distribution in ordinary, ilmenite and ilmenite-limonite concrete shields both along and perpendicular to the beam axis for both the direct reactor neutrons and the reactor thermal neutrons. (author).

  13. Swimming pool cleaner poisoning

    Science.gov (United States)

    Swimming pool cleaner poisoning occurs when someone swallows this type of cleaner, touches it, or breathes in ... The harmful substances in swimming pool cleaner are: Bromine ... copper Chlorine Soda ash Sodium bicarbonate Various mild acids

  14. Theoretical study for ICRF sustained LHD type p-{sup 11}B reactor

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Tsuguhiro (ed.)

    2003-04-01

    This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p-{sup 11}B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D-{sup 3}He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p-{sup 11}B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D-{sup 3}He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor ({gamma}{sub HH}). It is shown that high average beta plasma confinement, a large confinement factor ({gamma}{sub HH} > 3) and the hot ion mode (T{sub i}/T{sub e} > 1.4) are necessary to achieve the ignition of the D-{sup 3}He helical reactor. Characteristics of ICRF sustained p-{sup 11}B reactor are analyzed in section 4. The nuclear fusion reaction rate < {sigma}{upsilon} > is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p-{sup 11}B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

  15. Photosystem II Activity of Wild Type Synechocystis PCC 6803 and Its Mutants with Different Plastoquinone Pool Redox States.

    Science.gov (United States)

    Voloshina, O V; Bolychevtseva, Y V; Kuzminov, F I; Gorbunov, M Y; Elanskaya, I V; Fadeev, V V

    2016-08-01

    To assess the role of redox state of photosystem II (PSII) acceptor side electron carriers in PSII photochemical activity, we studied sub-millisecond fluorescence kinetics of the wild type Synechocystis PCC 6803 and its mutants with natural variability in the redox state of the plastoquinone (PQ) pool. In cyanobacteria, dark adaptation tends to reduce PQ pool and induce a shift of the cyanobacterial photosynthetic apparatus to State 2, whereas illumination oxidizes PQ pool, leading to State 1 (Mullineaux, C. W., and Holzwarth, A. R. (1990) FEBS Lett., 260, 245-248). We show here that dark-adapted Ox(-) mutant with naturally reduced PQ is characterized by slower QA(-) reoxidation and O2 evolution rates, as well as lower quantum yield of PSII primary photochemical reactions (Fv/Fm) as compared to the wild type and SDH(-) mutant, in which the PQ pool remains oxidized in the dark. These results indicate a large portion of photochemically inactive PSII reaction centers in the Ox(-) mutant after dark adaptation. While light adaptation increases Fv/Fm in all tested strains, indicating PSII activation, by far the greatest increase in Fv/Fm and O2 evolution rates is observed in the Ox(-) mutant. Continuous illumination of Ox(-) mutant cells with low-intensity blue light, that accelerates QA(-) reoxidation, also increases Fv/Fm and PSII functional absorption cross-section (590 nm); this effect is almost absent in the wild type and SDH(-) mutant. We believe that these changes are caused by the reorganization of the photosynthetic apparatus during transition from State 2 to State 1. We propose that two processes affect the PSII activity during changes of light conditions: 1) reversible inactivation of PSII, which is associated with the reduction of electron carriers on the PSII acceptor side in the dark, and 2) PSII activation under low light related to the increase in functional absorption cross-section at 590 nm.

  16. Evaluation of the aptitude for the service of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico; Evaluacion de la aptitud para el servicio de la piscina del reactor TRIGA Mark III del Instituto Nacional de Investigaciones Nucleares de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, J.; Gachuz M, M.; Diaz S, A.; Arganis J, C.; Gonzalez R, C.; Nava G, T.; Medina R, M.J. [Departamento de Sintesis y Caracterizacion de Materiales del ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    This work describes the evaluation of the structural integrity of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico, which was realized in July 2001, as an element to determine those actions for preventive and corrective maintenance which owner must do it for a safety and efficient operation of the component in the next years. (Author)

  17. Study of different types of mycobacteria in sediments of fish breeding pools of north of Iran

    Directory of Open Access Journals (Sweden)

    Ghazesaeed K

    1997-07-01

    Full Text Available In this study, 307 samples of the sediments of fish breeding pools of the different parts of North of Iran were tested for the survey of different environmental Mycobacteria. After the process of cultivation, 107 cases of Mycobacterium were gained which after the performance of different biochemical tests. 112 cases of Mycobacterium were identified. From among the isolated Mycobacteria, the highest rank belonged to M.fortuitum with the frequency of 13.97% and the next M.gordonae 10.66% M.xenopi, M.nonchromagenicum 8.2% and the last M.marinum with the frequency of 5.74%. M.marinum was the case of Tuberculosis of fish and had important role in the creation of granuloma. Next to that, M.fortuitum, M.kanasasii and M.gordenae had less importance. The existence of such Mycobacteria in the fish breeding pools were on one hand the cause of pollution of fish and on the other hand the fishman and other people who are somehow connected to the fish and the pools sediments are subject to disease in case of existence of injury in their hands or feet

  18. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the si

  19. NMR microimaging of fluid flow in model string-type reactors

    NARCIS (Netherlands)

    Koptyug, I.V.; Kovtunov, K.V.; Gerkema, E.; Kiwi-Minskerc, L.; Sagdeev, R.Z.

    2007-01-01

    Magnetic resonance microimaging (MRM) was employed to obtain quantitative velocity maps of water flowing in the channels possessing unconventional cross-section shapes formed by a bundle of parallel fibers within a tubular string-type reactor. The maps obtained demonstrate the presence of large amou

  20. Effects of reactor type and mass transfer on the morphology of CuS and ZnS crystals

    NARCIS (Netherlands)

    Al-Tarazi, Mousa; Heesink, A. Bert M.; Versteeg, Geert F.

    2005-01-01

    For the precipitation of CuS and ZnS, the effects of the reactor/precipitator type, mass transfer and process conditions on crystal morphology were studied. Either H2S gas or a S2- solution were applied. Three different types of reactors have been tested, namely a laminar jet, a bubble column and an

  1. Pool spacing in forest channels

    Science.gov (United States)

    David R. Montgomery; John M. Buffington; Richard D. Smith; Kevin M. Schmidt; George Pess

    1995-01-01

    Field surveys of stream channels in forested mountain drainage basins in southeast Alaska and Washington reveal that pool spacing depends on large woody debris (LWD) loading and channel type, slope, and width. Mean pool spacing in pool-riffle, plane-bed, and forced pool-riffle channels systematically decreases from greater than 13 channel widths per pool to less than 1...

  2. Coccomyxa actinabiotis sp. nov. (Trebouxiophyceae, Chlorophyta), a new green microalga living in the spent fuel cooling pool of a nuclear reactor.

    Science.gov (United States)

    Rivasseau, Corinne; Farhi, Emmanuel; Compagnon, Estelle; de Gouvion Saint Cyr, Diane; van Lis, Robert; Falconet, Denis; Kuntz, Marcel; Atteia, Ariane; Couté, Alain

    2016-10-01

    Life can thrive in extreme environments where inhospitable conditions prevail. Organisms which resist, for example, acidity, pressure, low or high temperature, have been found in harsh environments. Most of them are bacteria and archaea. The bacterium Deinococcus radiodurans is considered to be a champion among all living organisms, surviving extreme ionizing radiation levels. We have discovered a new extremophile eukaryotic organism that possesses a resistance to ionizing radiations similar to that of D. radiodurans. This microorganism, an autotrophic freshwater green microalga, lives in a peculiar environment, namely the cooling pool of a nuclear reactor containing spent nuclear fuels, where it is continuously submitted to nutritive, metallic, and radiative stress. We investigated its morphology and its ultrastructure by light, fluorescence and electron microscopy as well as its biochemical properties. Its resistance to UV and gamma radiation was assessed. When submitted to different dose rates of the order of some tens of mGy · h(-1) to several thousands of Gy · h(-1) , the microalga revealed to be able to survive intense gamma-rays irradiation, up to 2,000 times the dose lethal to human. The nuclear genome region spanning the genes for small subunit ribosomal RNA-Internal Transcribed Spacer (ITS) 1-5.8S rRNA-ITS2-28S rRNA (beginning) was sequenced (4,065 bp). The phylogenetic position of the microalga was inferred from the 18S rRNA gene. All the revealed characteristics make the alga a new species of the genus Coccomyxa in the class Trebouxiophyceae, which we name Coccomyxa actinabiotis sp. nov.

  3. Laboratory data for review of outlet water temperature limits for BDF type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.; Fitzsimmons, D.E.

    1960-12-11

    A knowledge of the thermal and hydraulic conditions within a reactor fuel channel during an inadvertent flow reduction is needed to establish reactor operating limits. Such limits, which are based on outlet water temperature, are called ``trip-after-instability`` limits by the reactor operating personnel. Laboratory experiments were performed to update the knowledge of such conditions in a BDF reactor type fuel channel while using internally and externally cooled fuel elements (I&E`s) at tube powers up to 1530 KW. In addition to a general extension of previous data, the new data were used to review certain specific details involved in ``trip-after-instability`` limit calculations. It was found that in calculating the limits, the isothermal pressure drop across the fuel elements must be related to flow rate by the exponent 1.8, ({delta}P {proportional_to} F{sup 1.8}), rather than the more convenient value of 2.0. It was found that this method of limit determination is applicable to the high rear header pressures presently attained on the reactors and also applicable to tubes with very low Panellit pressures. And finally, the validity of certain analytical transformations of experimental data, called generalization of hydraulic demand curves, was reaffirmed for the above conditions.

  4. SVBR-100 module-type fast reactor of the IV generation for regional power industry

    Science.gov (United States)

    Zrodnikov, A. V.; Toshinsky, G. I.; Komlev, O. G.; Stepanov, V. S.; Klimov, N. N.

    2011-08-01

    In the report the following is presented: basic conceptual provisions of the innovative nuclear power technology (NPT) based on modular fast reactors (FR) SVBR-100, summarized results of calculations of the reactor, analysis of the opportunities of multi-purpose application of such reactor facilities (RF) including export potentials with due account of nonproliferation requirements. The most important features of the proposed NPT analyzed in the report are as follows: (1) integral (monoblock) arrangement of the primary circuit equipment with entire elimination of the primary circuit pipelines and valves that considerably reduces the construction and assembly works period and coupling with high boiling point of lead-bismuth coolant (LBC) deterministically eliminates accidents of the LOCA type, (2) option for 100 MWe power and dimensions of the reactor provide: on the one hand, an opportunity to transport the reactor monoblock in factory-readiness by railway as well as other kinds of transport, on the other hand, core breeding ratio (CBR) exceeds 1 while MOX-fuel is used. The preferable area of application of RF SVBR-100 is regional and small power requiring power-units of electric power in a range of (100-600) MW, which could be used for cogeneration-based district heating while locating them nearby cities as well as for generation of electric power in a mode of load tracking in the regions with low network systems.

  5. Power regulating range broadening of the WWER-type reactor power units

    Energy Technology Data Exchange (ETDEWEB)

    Dement' ev, B.A.; Petrov, V.A.; Proskuryakov, A.G.; Puchkov, V.V. (Moskovskij Ehnergeticheskij Inst. (USSR))

    1984-02-01

    Calculational studies on the use of sliding pressure (SP) regime to expand the regulating range of the WWER-440 reactor power units are presented. Two operation regimes of a power unit have been considered: according to weekly and daily load swings in electrical grids. The conclusion is made that the use of SP regime in the secondary circuit improves manoeuvable characterstics of the power unit in the second half of operating cycle. T of the reactor (0.6 reactor. Besides, the use of SP regime during power unit operation with decreased loadings is the more efficient the smaller is the load. The range of operating cycle 0.8 <= T <= 1 makes the greatest contribution to regulating range broadening as a result of SP regime use. Conclusions of the calculational studies can be also applied to WWER reactors of other types as well as to RBMK reactors.

  6. Atomistic simulation on charge mobility of amorphous tris(8-hydroxyquinoline) aluminum (Alq3): origin of Poole-Frenkel-type behavior.

    Science.gov (United States)

    Nagata, Yuki; Lennartz, Christian

    2008-07-21

    The atomistic simulation of charge transfer process for an amorphous Alq(3) system is reported. By employing electrostatic potential charges, we calculate site energies and find that the standard deviation of site energy distribution is about twice as large as predicted in previous research. The charge mobility is calculated via the Miller-Abrahams formalism and the master equation approach. We find that the wide site energy distribution governs Poole-Frenkel-type behavior of charge mobility against electric field, while the spatially correlated site energy is not a dominant mechanism of Poole-Frenkel behavior in the range from 2x10(5) to 1.4x10(6) V/cm. Also we reveal that randomly meshed connectivities are, in principle, required to account for the Poole-Frenkel mechanism. Charge carriers find a zigzag pathway at low electric field, while they find a straight pathway along electric field when a high electric field is applied. In the space-charge-limited current scheme, the charge-carrier density increases with electric field strength so that the nonlinear behavior of charge mobility is enhanced through the strong charge-carrier density dependence of charge mobility.

  7. The use of experimental data in an MTR-type nuclear reactor safety analysis

    Science.gov (United States)

    Day, Simon E.

    Reactivity initiated accidents (RIAs) are a category of events required for research reactor safety analysis. A subset of this is unprotected RIAs in which mechanical systems or human intervention are not credited in the response of the system. Light-water cooled and moderated MTR-type ( i.e., aluminum-clad uranium plate fuel) reactors are self-limiting up to some reactivity insertion limit beyond which fuel damage occurs. This characteristic was studied in the Borax and Spert reactor tests of the 1950s and 1960s in the USA. This thesis considers the use of this experimental data in generic MTR-type reactor safety analysis. The approach presented herein is based on fundamental phenomenological understanding and uses correlations in the reactor test data with suitable account taken for differences in important system parameters. Specifically, a semi-empirical approach is used to quantify the relationship between the power, energy and temperature rise response of the system as well as parametric dependencies on void coefficient and the degree of subcooling. Secondary effects including the dependence on coolant flow are also examined. A rigorous curve fitting approach and error assessment is used to quantify the trends in the experimental data. In addition to the initial power burst stage of an unprotected transient, the longer term stability of the system is considered with a stylized treatment of characteristic power/temperature oscillations (chugging). A bridge from the HEU-based experimental data to the LEU fuel cycle is assessed and outlined based on existing simulation results presented in the literature. A cell-model based parametric study is included. The results are used to construct a practical safety analysis methodology for determining reactivity insertion safety limits for a light-water moderated and cooled MTR-type core.

  8. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L.; Camacho L, M.E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  9. Investigating signs of recent evolution in the pool of proviral HIV type 1 DNA during years of successful HAART

    DEFF Research Database (Denmark)

    Mens, Helene; Pedersen, Anders G; Jørgensen, Louise B

    2007-01-01

    In order to shed light on the nature of the persistent reservoir of human immunodeficiency virus type 1 (HIV-1), we investigated signs of recent evolution in the pool of proviral DNA in patients on successful HAART. Pro-viral DNA, corresponding to the C2-V3-C3 region of the HIV-1 env gene......, was collected from PBMCs isolated from 57 patients. Both "consensus" (57 patients) and clonal (7 patients) sequences were obtained from five time points spanning a 24-month period. The main computational strategy was to use maximum likelihood to fit a set of alternative phylogenetic models to the clonal data...

  10. Investigating signs of recent evolution in the pool of proviral HIV type 1 DNA during years of successful HAART

    DEFF Research Database (Denmark)

    Mens, Helene; Pedersen, Anders G; Jørgensen, Louise B;

    2007-01-01

    In order to shed light on the nature of the persistent reservoir of human immunodeficiency virus type 1 (HIV-1), we investigated signs of recent evolution in the pool of proviral DNA in patients on successful HAART. Pro-viral DNA, corresponding to the C2-V3-C3 region of the HIV-1 env gene......, and then determine the support for models that imply evolution between time points. Model fit and model-selection uncertainty was assessed using the Akaike information criterion (AIC) and Akaike weights. The consensus sequence data was also analyzed using a range of phylogenetic techniques to determine whether...

  11. Pool scrubbing models for iodine components

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, K. [Battelle Ingenieurtechnik GmbH, Eschborn (Germany)

    1996-12-01

    Pool scrubbing is an important mechanism to retain radioactive fission products from being carried into the containment atmosphere or into the secondary piping system. A number of models and computer codes has been developed to predict the retention of aerosols and fission product vapours that are released from the core and injected into water pools of BWR and PWR type reactors during severe accidents. Important codes in this field are BUSCA, SPARC and SUPRA. The present paper summarizes the models for scrubbing of gaseous Iodine components in these codes, discusses the experimental validation, and gives an assessment of the state of knowledge reached and the open questions which persist. The retention of gaseous Iodine components is modelled by the various codes in a very heterogeneous manner. Differences show up in the chemical species considered, the treatment of mass transfer boundary layers on the gaseous and liquid sides, the gas-liquid interface geometry, calculation of equilibrium concentrations and numerical procedures. Especially important is the determination of the pool water pH value. This value is affected by basic aerosols deposited in the water, e.g. Cesium and Rubidium compounds. A consistent model requires a mass balance of these compounds in the pool, thus effectively coupling the pool scrubbing phenomena of aerosols and gaseous Iodine species. Since the water pool conditions are also affected by drainage flow of condensate water from different regions in the containment, and desorption of dissolved gases on the pool surface is determined by the gas concentrations above the pool, some basic limitations of specialized pool scrubbing codes are given. The paper draws conclusions about the necessity of coupling between containment thermal-hydraulics and pool scrubbing models, and proposes ways of further simulation model development in order to improve source term predictions. (author) 2 tabs., refs.

  12. Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

    Science.gov (United States)

    Doshi, Bharat; Reddy, D. Chenna

    2017-04-01

    Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

  13. Study on Thermal-Hydraulic Behavior of an Integral Type Reactor under Heaving Condition

    Directory of Open Access Journals (Sweden)

    Beibei Feng

    2014-01-01

    Full Text Available A self-developed program was used to study the thermal-hydraulic behavior of an integral type reactor under heaving condition. Comparison of calculated results with the data of experiments performed on a natural circulation loop designed with reference to an integral type reactor of Tsinghua University in inclination, heaving, and rolling motions was carried out. Characteristics of natural circulation in heaving motion and effect of motion parameters on natural circulation were investigated. Results indicated that: (1 long-period heaving motion would lead to more significant influence than inclination and rolling motion; (2 it was an alternating force field which consisted of gravity and an additional force that decided the flow temperature and density difference of natural circulation; (3 effect of strength k and cycle T of heaving motion on flow fluctuation of natural circulation and condensate depression of heating section outlet was performed.

  14. Hydrogen energy recovery from high strength organic wastewater with ethanol type fermentation using acidogenic EGSB reactor

    Institute of Scientific and Technical Information of China (English)

    REN Nan-qi; GUO Wan-qian; WANG Xiang-jing; ZHANG Lu-si

    2005-01-01

    A lab-scale expanded granular sludge bed (EGSB) reactor was employed to evaluate the feasibility of the hydrogen energy recovery potential from high strength organic wastewater. The results showed that a maxioperation. At the acidogenic phase, COD removal rate was stable at about 15%. In the steady operation peri od, the main liquid end products were ethanol and acetic acid, which represented ethanol type fermentation. Among the liquid end products, the concentration percentage of ethanol and acetic acid amounted to 69.5% ~89.8% and the concentration percentage of ethanol took prominent about 51.7% ~ 59.1%, which is better than the utilization of substrate for the methanogenic bacteria. An ethanol type fermentation pathway was suggested in the operation of enlarged industrial continuous hydrogen bio-producing reactors.

  15. Deployable nuclear fleet based on available quantities of uranium and reactor types – the case of fast reactors started up with enriched uranium

    Directory of Open Access Journals (Sweden)

    Baschwitz Anne

    2016-01-01

    Full Text Available International organizations regularly produce global energy demand scenarios. To account for the increasing population and GDP trends, as well as to encompass evolving energy uses while satisfying constraints on greenhouse gas emissions, long-term installed nuclear power capacity scenarios tend to be more ambitious, even after the Fukushima accident. Thus, the amounts of uranium or plutonium needed to deploy such capacities could be limiting factors. This study first considers light-water reactors (LWR, GEN III using enriched uranium, like most of the current reactor technologies. It then examines the contribution of future fast reactors (FR, GEN IV operating with an initial fissile load and then using depleted uranium and recycling their own plutonium. However, as plutonium is only available in limited quantity since it is only produced in nuclear reactors, the possibility of starting up these Generation IV reactors with a fissile load of enriched uranium is also explored. In one of our previous studies, the uranium consumption of a third-generation reactor like an EPR™ was compared with that of a fast reactor started up with enriched uranium (U5-FR. For a reactor lifespan of 60 years, the U5-FR consumes three times less uranium than the EPR and represents a 60% reduction in terms of separative work units (SWU, though its requirements are concentrated over the first few years of operation. The purpose of this study is to investigate the relevance of U5-FRs in a nuclear fleet deployment configuration. Considering several power demand scenarios and assuming different finite quantities of available natural uranium, this paper examines what types of reactors must be deployed to meet the demand. The deployment of light-water reactors only is not sustainable in the long run. Generation IV reactors are therefore essential. Yet when started up with plutonium, the number of reactors that can be deployed is also limited. In a fleet deployment

  16. A High Operability Supervisory Digital System for TRIGA-Type Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aronica, O.; Bove, R.; Cappelli, M.; Falconi, L.; Palomba, M.; Santoro, E.; Sepielli, M. [ENEA, UTFISST, Casaccia Research Center, Via Anguillarese, 301 Rome (Italy); Memmi, F. [University of Rome ' Roma Tre' , Department of Electrical Engineering, Via della Vasca Navale, 84 Rome (Italy)

    2011-07-01

    In this work, we propose an outline of a monitoring system to supervise variables coming from a fission nuclear reactor of TRIGA type (1-MW TRIGA reactor RC-1). The system can interface the control room instrumentation and can display the characteristic parameters (e.g. nuclear power, temperatures, flow rates, radiological parameters) in an intuitive, user-friendly way for plant operators. This aim is achieved using the Labview development environment. A front panel of a virtual instrument allows for a direct measure and a check that would not be possible by only reading the output data coming from the instruments of the control room, because of their standards and strict safety regulations. The acquisition system, for signals coming from the reactor, can process data and generate a detailed representation of the results. Statistics resulting from data analysis will be interpreted to optimize reactor management parameters. This system also includes a simulation tool to predict specific performances and investigate critical phenomena, or to optimize overall plant performances. In particular, it allows to have a feedback control and to perform predictive statistical surveys of all main process parameters. (author)

  17. Dismantling design for a reference research reactor of the WWR type

    Energy Technology Data Exchange (ETDEWEB)

    Lobach, Yu.N., E-mail: lobach@kinr.kiev.ua [Institute for Nuclear Research, Pr. Nauki, 47, Kiev 03680 (Ukraine); Cross, M.T., E-mail: Martin.Cross@nuvia.co.uk [Nuvia Ltd., Robinson House, Crow Park Way, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HY (United Kingdom)

    2014-01-15

    Highlights: • Design features of WWRs relevant to decommissioning have been analysed. • The technical basis for the preparation and implementation of dismantling has been established for a reference WWR. • The applicability of existing proven dismantling technologies has been established. -- Abstract: A decommissioning study has been carried out for a reference research reactor of the WWR type. Many such reactors were constructed more than 50 years ago and most of them are still in operation. Decommissioning has now become an important consideration. This paper summarizes the main decommissioning steps and, on the basis of the reactor design features, technical aspects of the dismantling and removal of the contaminated/activated components have been analysed. The advisability of the removal of large components, such as the reactor vessel and the heat-exchangers, as one piece items has also been demonstrated. Additionally, a work schedule and an estimation of the collective dose for the preparation and implementation of dismantling have been established. The applicability of existing proven dismantling technologies has been identified together with some additional features for the dismantling.

  18. A Study of the Energy Efficiency of Hadronic Reactors of Molecular Type

    OpenAIRE

    Aringazin, A. K.; Santilli, R. M.

    2001-01-01

    In this paper, we introduce an estimate of the "commercial efficiency" of Santilli's hadronic reactors of molecular type (Patented and International Patents Pending) which convert a liquid feedstock (such as automotive antifreeze and oil waste, city or farm liquid waste, crude oil, etc.) into the clean burning magnegas plus heat acquired by the liquid feedstock. The "commercial efficiency" is defined as the ratio between the total energy output (energy in magnegas plus heat) and the electric ...

  19. Performance of static var compensator control type thyristor controlled reactor and thyristor switched capacitor

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Josias M. de; Yung, Chou Shaw; Rose, Eber H.; Pantoja, Antonio L.A. [ELETRONORTE, Belem, PA (Brazil); Fouesnant, Thomas; Boissier, Luc

    1994-12-31

    This paper has the objective of presenting the philosophy of Static Var Compensator (SVC) Control as well the necessary adjustments in the project of control system to guarantee suitable performance under different operating conditions. The verification on the performance of the SVC control has been done by Transient Network Analyzer (TNA/CEPEL) studies, commissioning tests and a factory tests. The SVC is the type of Thyristor Controlled Reactor (TCR) and Thyristor Switched Capacitor (TSC). (author) 3 refs., 12 figs.

  20. Synthesis of layered double hydroxide nanosheets by coprecipitation using a T-type microchannel reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pang, Xiujiang; Sun, Meiyu; Ma, Xiuming [State Key Laboratory Base of Eco-chemical Engineering, College of Chemistry and Molecular Engineering, Qingdao University of Science and Technology, Qingdao 266042 (China); Hou, Wanguo, E-mail: wghou@sdu.edu.cn [Key Laboratory of Colloid and Interface Chemistry (Ministry of Education), Shandong University, Jinan 250100 (China)

    2014-02-15

    The synthesis of Mg{sub 2}Al–NO{sub 3} layered double hydroxide (LDH) nanosheets by coprecipitation using a T-type microchannel reactor is reported. Aqueous LDH nanosheet dispersions were obtained. The LDH nanosheets were characterized by X-ray diffraction, transmission electron microscopy, atomic force microscopy and particle size analysis, and the transmittance and viscosity of LDH nanosheet dispersions were examined. The two-dimensional LDH nanosheets consisted of 1–2 brucite-like layers and were stable for ca. 16 h at room temperature. In addition, the co-assembly between LDH nanosheets and dodecyl sulfate (DS) anions was carried out, and a DS intercalated LDH nanohybrid was obtained. To the best of our knowledge, this is the first report of LDH nanosheets being directly prepared in bulk aqueous solution. This simple, cheap method can provide naked LDH nanosheets in high quantities, which can be used as building blocks for functional materials. - Graphical abstract: Layered double hydroxide (LDH) nanosheets were synthesized by coprecipitation using a T-type microchannel reactor, and could be used as basic building blocks for LDH-based functional materials. Display Omitted - Highlights: • LDH nanosheets were synthesized by coprecipitation using a T-type microchannel reactor. • Naked LDH nanosheets were dispersed in aqueous media. • LDH nanosheets can be used as building blocks for functional materials.

  1. LTP requires a reserve pool of glutamate receptors independent of subunit type.

    Science.gov (United States)

    Granger, Adam J; Shi, Yun; Lu, Wei; Cerpas, Manuel; Nicoll, Roger A

    2013-01-24

    Long-term potentiation (LTP) of synaptic transmission is thought to be an important cellular mechanism underlying memory formation. A widely accepted model posits that LTP requires the cytoplasmic carboxyl tail (C-tail) of the AMPA (α-amino-3-hydroxy-5-methyl-4-isoxazole propionic acid) receptor subunit GluA1. To find the minimum necessary requirement of the GluA1 C-tail for LTP in mouse CA1 hippocampal pyramidal neurons, we used a single-cell molecular replacement strategy to replace all endogenous AMPA receptors with transfected subunits. In contrast to the prevailing model, we found no requirement of the GluA1 C-tail for LTP. In fact, replacement with the GluA2 subunit showed normal LTP, as did an artificially expressed kainate receptor not normally found at these synapses. The only conditions under which LTP was impaired were those with markedly decreased AMPA receptor surface expression, indicating a requirement for a reserve pool of receptors. These results demonstrate the synapse's remarkable flexibility to potentiate with a variety of glutamate receptor subtypes, requiring a fundamental change in our thinking with regard to the core molecular events underlying synaptic plasticity.

  2. Types of dietary fat and breast cancer : a pooled analysis of cohort studies

    NARCIS (Netherlands)

    Smith-Warner, S.A.; Spiegelman, D.; Adami, H.O.; Beeson, W.L.; Brandt, P.A. van den; Folsom, A.R.; Fraser, G.E.; Freudenheim, J.L.; Goldbohm, R.A.; Graham, S.; Kushi, L.H.; Miller, A.B.; Rohan, T.E.; Speizer, F.E.; Toniolo, P.; Willett, W.C.; Wolk, A.; Zelenuch-Jacquotte, A.; Hunter, D.J.

    2001-01-01

    Recently, there has been interest in whether intakes of specific types of fat are associated with breast cancer risk independently of other types of fat, but results have been inconsistent. We identified 8 prospective studies that met predefined criteria and analyzed their primary data using a stand

  3. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  4. Development of core fuel management code system for WWER-type reactors

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered. The verification and validation of the code system have been examined through the IAEA WWER-1000-type Kalinin NPP benchmark problem. The numerical results are in good agreement with measurements and are close to those of other international institutes.

  5. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  6. PWR type reactors. Normal and accidental operation; Reacteurs a eau sous pression. Fonctionnement normal et accidentel

    Energy Technology Data Exchange (ETDEWEB)

    Petetrot, J.F. [AREVA NP, Dept. Fonctionnement Reacteur et Etudes d' Accidents/Division, Tour AREVA, 92 - Paris La Defense (France)

    2009-07-15

    This article presents the general operation principles of PWR type reactors with the limits to be respected for the core and the steam supply system. Regulation systems controlling the main parameters are described as well: measurements used, functional structures, controlled actuators. The protection system which can lead to the automatic shutdown of the reactor (emergency rod drop) and to the start-up of safeguard functions is detailed. The interface for the conventional protection system is briefly described. The operation of the steam supply system with respect to the power grid needs is presented in relation with the regulation of the turbogenerator set: load follow-up, primary and secondary adjustment. Finally, the changes of the most important parameters during typical transients are commented. The main operations needed to move from the cold shutdown state to the nominal power are described as well. (J.S.)

  7. TASS/SMR code improvement for small break LOCA applicability at an integral type reactor, SMART

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Young-Jong, E-mail: chung@kaeri.re.kr; Kim, Soo-Hyung; Lim, Sung-Won; Bae, Kyoo-Hwan

    2015-12-15

    Highlights: • SMART adopts a passive system to enhance its safety. • TASS/SMR code is developed to analyze thermal hydraulic phenomena of the SMART plant. • Improved TASS/SMR code predicts well the results of the OSU-MASLWR total-loss-of-feedwater test. - Abstract: Small reactors are a suitable option for nuclear system deployment in developing countries or non-electrical applications for various facilities. SMART is one of the small integral type reactors to apply flexibly local power demands or sea water desalination. A thermal hydraulic analysis code, TASS/SMR, having SMART specific models, was developed to simulate thermal hydraulic phenomena of the SMART plant. The improved TASS/SMR code predicts well the system behaviors under two-phase conditions compared with the OSU-MASLWR experimental results. A small break LOCA simulation of the SMART plant is improved a void distribution, a break flow, and a collapsed water level in the core.

  8. A Study of the Energy Efficiency of Hadronic Reactors of Molecular Type

    CERN Document Server

    Aringazin, A K

    2001-01-01

    In this paper, we introduce an estimate of the "commercial efficiency" of Santilli's hadronic reactors of molecular type (Patented and International Patents Pending) which convert a liquid feedstock (such as automotive antifreeze and oil waste, city or farm liquid waste, crude oil, etc.) into the clean burning magnegas plus heat acquired by the liquid feedstock. The "commercial efficiency" is defined as the ratio between the total energy output (energy in magnegas plus heat) and the electric energy used for its production, while the "scientific efficiency" is the usual ratio between the total energy output and the total energy input (the sum of the electric energy plus the energy in the liquid feedstock as well as that in the carbon electrodes). A primary purpose of this paper is to show that conventional thermochemistry does indeed predict a commercial efficiency bigger than one, although their values is considerably smaller than the actual efficiency measured in the reactors, thus indicating the applicabili...

  9. Assessing optimal fermentation type for bio-hydrogen production in continuous-flow acidogenic reactors.

    Science.gov (United States)

    Ren, N Q; Chua, H; Chan, S Y; Tsang, Y F; Wang, Y J; Sin, N

    2007-07-01

    In this study, the optimal fermentation type and the operating conditions of anaerobic process in continuous-flow acidogenic reactors was investigated for the maximization of bio-hydrogen production using mixed cultures. Butyric acid type fermentation occurred at pH>6, propionic acid type fermentation occurred at pH about 5.5 with E(h) (redox potential) >-278mV, and ethanol-type fermentation occurred at pHhydrogen production capacities between the fermentation types, which remained stable when the organic loading rate (OLR) reached the highest OLR at 86.1kgCOD/m(3)d. The maximum hydrogen production reached up to 14.99L/d.

  10. Evaluation of Total Coliform, Fecal Coliform and Residual Chlorine in Swimming Pools in Kermanshah on the Season, the type of Pool, Disinfection System and Source of Water Supply in the during of three years (2010-2012

    Directory of Open Access Journals (Sweden)

    K SHarafi

    2014-11-01

    From the results , although the pools of water quality parameters has been studied in almost ideal But in summer, especially on a female pools and pools with wells water supply source than other pools , to be more oversight .

  11. Maternal age at birth and childhood type 1 diabetes: a pooled analysis of 30 observational studies

    DEFF Research Database (Denmark)

    Cardwell, Chris R; Stene, Lars C; Joner, Geir

    2009-01-01

    for potential confounders. Meta-analysis techniques were used to derive combined odds ratios and to investigate heterogeneity among studies. RESULTS: Data were available for 5 cohort and 25 case-control studies, including 14,724 cases of type 1 diabetes. Overall, there was, on average, a 5% (95% CI 2...

  12. Investigating signs of recent evolution in the pool of proviral HIV type 1 DNA during years of successful HAART

    DEFF Research Database (Denmark)

    Mens, Helene; Pedersen, Anders G; Jørgensen, Louise B

    2007-01-01

    In order to shed light on the nature of the persistent reservoir of human immunodeficiency virus type 1 (HIV-1), we investigated signs of recent evolution in the pool of proviral DNA in patients on successful HAART. Pro-viral DNA, corresponding to the C2-V3-C3 region of the HIV-1 env gene......, and then determine the support for models that imply evolution between time points. Model fit and model-selection uncertainty was assessed using the Akaike information criterion (AIC) and Akaike weights. The consensus sequence data was also analyzed using a range of phylogenetic techniques to determine whether...... there were temporal trends indicating ongoing replication and evolution. In summary, it was not possible to detect definitive signs of ongoing evolution in either the bulk-sequenced or the clonal data with the methods employed here, but our results could be consistent with localized expression of archival...

  13. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  14. On the Scale-up of Gas-Hydrate-Forming Reactors: The Case of Gas-Dispersion-Type Reactors

    Directory of Open Access Journals (Sweden)

    Yasuhiko H. Mori

    2015-02-01

    Full Text Available For establishing hydrate-based technologies for natural-gas storage/transport, CO2 capture from industrial flue gases, etc., we need appropriate guidelines for the scale-up of hydrate production/processing equipment from laboratory scales to industrial scales. This paper aims to provide technical remarks on the scale-up of hydrate-forming reactors, the central components of hydrate production/processing equipment, particularly focusing on such a reactor design that hydrate-forming gas is dispersed in an aqueous phase which is either stirred in a tank or forced to flow through a tube. Based on the principles of classical fluid mechanics and heat-transfer analysis, the paper derives semi-empirical formulas that show how the capacity for heat discharge from each reactor and the power for operating the reactor are required to change with an increase in its size. Consequently, it is concluded that the stirred-tank design is unfavorable for significant scale-up and that the scale-up of tubular reactors should be made without significantly increasing the in-tube flow velocity.

  15. Improvement of nuclear ship engineering simulation system. Hardware renewal and interface improvement of the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroki; Kyoya, Masahiko; Shimazaki, Junya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kano, Tadashi [KCS, Co., Mito, Ibaraki (Japan); Takahashi, Teruo [Energis, Co., Kobe, Hyogo (Japan)

    2001-10-01

    JAERI had carried out the design study about a lightweight and compact integral type reactor (an advanced marine reactor) with passive safety equipment as a power source for the future nuclear ships, and completed an engineering design. We have developed the simulator for the integral type reactor to confirm the design and operation performance and to utilize the study of automation of the reactor operation. The simulator can be used also for future research and development of a compact reactor. However, the improvement in a performance of hardware and a human machine interface of software of the simulator were needed for future research and development. Therefore, renewal of hardware and improvement of software have been conducted. The operability of the integral-reactor simulator has been improved. Furthermore, this improvement with the hardware and software on the market brought about better versatility, maintainability, extendibility and transfer of the system. This report mainly focuses on contents of the enhancement in a human machine interface, and describes hardware renewal and the interface improvement of the integral type reactor simulator. (author)

  16. Artificial intelligence applied to fuel management in BWR type reactors; Inteligencia artificial aplicada a la administracion de combustible en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz S, J.J

    1998-10-01

    In this work two techniques of artificial intelligence, neural networks and genetic algorithms were applied to a practical problem of nuclear fuel management; the determination of the optimal fuel reload for a BWR type reactor. This is an important problem in the design of the operation cycle of the reactor. As a result of the application of these techniques, comparable or even better reloads proposals than those given by expert companies in the subject were obtained. Additionally, two other simpler problems in reactor physics were solved: the determination of the axial power profile and the prediction of the value of some variables of interest at the end of the operation cycle of the reactor. Neural networks and genetic algorithms have been applied to solve many problems of engineering because of their versatility but they have been rarely used in the area of fuel management. The results obtained in this thesis indicates the convenience of undertaking further work on this area and suggest the application of these techniques of artificial intelligence to the solution of other problems in nuclear reactor physics. (Author)

  17. Abundance and Dynamics of Soil Labile Carbon Pools Under Different Types of Forest Vegetation

    Institute of Scientific and Technical Information of China (English)

    JIANG Pei-Kun; XU Qiu-Fang

    2006-01-01

    Soil organic matter (SOM) in forest ecosystems is not only important to global carbon (C) storage but also to sustainable management of forestland with vegetation types, being a critical factor in controlling the quantity and dynamics of SOM. In this field experiment soil plots with three replicates were selected from three forest vegetation types: broadleaf,Masson pine (Pinus massoniana Lamb.), and Chinese fir (Cunninghamia lanceolata Hook.). Soil total organic C (TOC),two easily oxidizable C levels (EOC1 and EOC2, which were oxidized by 66.7 mmol L-1 K2Cr2O7 at 130-140 ℃ and333 mmol L-1 KMnO4 at 25 ℃, respectively), microbial biomass C (MBC), and water-soluble organic C (WSOC)were analyzed for soil samples. Soil under the broadleaf forest stored significantly higher TOC (P ≤ 0.05). Because of its significantly larger total soil C storage, the soil under the broadleaf forest usually had significantly higher levels (P ≤ 0.05)of the different labile organic carbons, EOC1, EOC2, MBC, and WSOC; but when calculated as a percentage of TOC each labile C fraction of the broadleaf forest was significantly lower (P ≤ 0.05) than one of the other two forests. Under all the three vegetation types temperature as well as quality and season of litter input generally affected the dynamics of different organic C fractions in soils, with EOC1, EOC2, and MBC increasing closely following increase in temperature,whereas WSOC showed an opposite trend.

  18. MR-6 type fuel elements cooling in natural convection conditions after the reactor shut down

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Bykowski, W.; Moldysz, A. [Institute of Atomic Energy, Otwock Swierk (Poland)

    2002-07-01

    Natural cooling conditions of the nuclear fuel in the channel type reactor after its shut down are commonly determined with relatively high uncertainty. This is not only to he lack of adequate measurements of thermal parameters i.e. the residual power generation, the coolant flow and temperatures, but also due to indeterminate model of convection mechanism. The numerical simulation of natural convection in multitube fuel assembly in the fuel channel leads to various convection modes including evidently chaotic behaviour. To determine the real cooling conditions in the MARIA research reactor a series of experiments has been performed with fuel assembly equipped with a set of thermocouples. After some forced cooling period (the shortest was half an hour after the reactor shut down) the reactor was left with the only natural convection. Two completely different cooling modes have been observed. The MARIA core consists of series of individual fuel channel and so called bypasses, maintaining the hydraulic properties of the fuel channel, connected in parallel. Initially, the convection cells were established trough few so-called bypasses providing a very effective mode of cooling. In this mode the flow charts were identical to those existing in forced cooling mode. After certain period the system switched on the second cooling mode with natural circulation within the individual fuel cells. Higher temperatures and temperature fluctuations were characteristic for this mode approaching 30 deg in amplitude. In almost all the cases the system was switching few times between modes, but eventually remained in the second mode. The switching times were not regular and the process has a chaotic behaviour. (author)

  19. Chemistry aspects of the source term formation for a severe accident in a CANDU type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Constantin, A.; Constantin, M. [Institute for Nuclear Research, Pitesti (Romania)

    2013-07-15

    The progression of a severe accident in a CANDU type reactor is slow because the core is surrounded by a large quantity of heavy and light water which acts as a heat sink to remove the decay heat. Therefore, the source term formation is a complex and long process involving fission products transport and releasing in the fuel matrix, thermal hydraulics of the transport fluid in the primary heat system and containment, deposition and transport of fission products, chemistry including the interaction with the dousing system, structural materials and paints, etc. The source term is strongly dependent on initial conditions and accident type. The paper presents chemistry aspects for a severe accident in a CANDU type reactor, in terms of the retention in the primary heat system. After releasing from the fuel elements, the fission products suffer a multitude of phenomena before they are partly transferred into the containment region. The most important species involved in the deposition were identified. At the same time, the influence of the break position in the transfer fractions from the primary heat system to the containment was investigated. (orig.)

  20. Proposal of rectifier type superconducting fault current limiter with non-inductive reactor (SFCL)

    Science.gov (United States)

    Mohammad Salim, Khosru; Muta, Itsuya; Hoshino, Tsutomu; Nakamura, Taketsune; Yamada, Masato

    2004-03-01

    A rectifier type superconducting fault current limiter (SFCL) with non-inductive reactor has been proposed. The concept behind this SFCL is the appearance of high impedance during non-superconducting state of the coil. In a hybrid bridge circuit, two superconducting coils connected in anti-parallel: a trigger coil and a limiting coil. Both the coils are magnetically coupled with each other and have same number of turns. There is almost zero flux inside the core and therefore the total inductance is small during normal operation. At fault time when the trigger coil current reaches to a certain level, the trigger coil changes from superconducting state to normal state. This super-to-normal transition of the trigger coil changes the current ratio of the coils and therefore the flux inside the reactor is no longer zero. So, the equivalent impedance of both the coils increased thus limits the fault current. We have carried out computer simulation using EMTDC and observed the results. A preliminary experiment has already been performed using copper wired reactor with simulated super-to-normal transition resistance and magnetic switches. Both the simulation and preliminary experiment shows good results. The advantage of using hybrid bridge circuit is that the SFCL can also be used as circuit breaker. Two separate bridge circuit can be used for both trigger coil and the limiter coil. In such a case, the trigger coil can be shutdown immediately after the fault to reduce heat and thus reduce the recovery time. Again, at the end of fault when the SFCL needs to re-enter to the grid, turning off the trigger circuit in the two-bridge configuration the inrush current can be reduced. This is because the current only flows through the limiting coil. Another advantage of this type of SFCL is that no voltage sag will appear during load increasing time as long as the load current stays below the trigger current level.

  1. Experimental and field approach to the hydraulics of nature-like pool-type fish migration facilities

    Directory of Open Access Journals (Sweden)

    Wang R.W.

    2011-02-01

    Full Text Available Nature-like fish migration facilities have gradually become a common type to ensure longitudinal connectivity of fish movements in running waters. This article presents verification on hydraulic and geometric parameters of nature-like pool-type fish passes via experimental and field investigations. The experiment verified that the maximum streamwise velocity near a slot ranged from 0.8–1.0 time of the theoretical maximum velocity. Large vertical recirculations presented below sills, moved downstream with the increase in discharge, and were likely to vanish or to change the rotation direction with high flow conditions. High turbulent kinetic energy distributed immediately downstream from boulder sills instead of along the water jet. Fieldwork was conducted at a full-width ramp in Kolbermoor and a partial-width ramp in Leitner in Bavaria under low, mean and high flow conditions to investigate the flow and geometry characteristics in real constructions and under various hydrologic conditions. The results for velocity show confidence in the method to obtain the maximum value around a slot by measuring at one depth only. Instead of flow velocity, water depth played a more critical role in the performance of a nature-like fishway, in particular under low flow conditions, and so did the arrangement of boulders along a sill. A detailed hydraulic/geometric investigation, together with biological monitoring, should be conducted to identify appropriate criteria on assessment of fish free passage at nature-like fish migration facilities.

  2. Issues of intergranular embrittlement of VVER-type nuclear reactors pressure vessel materials

    Science.gov (United States)

    Zabusov, O.

    2016-04-01

    In light of worldwide tendency to extension of service life of operating nuclear power plants - VVER-type in the first place - recently a special attention is concentrated on phenomena taking place in reactor pressure vessel materials that are able to lead to increased level of mechanical characteristics degradation (resistibility to brittle fracture) during long term of operation. Formerly the hardening mechanism of degradation (increase in the yield strength under influence of irradiation) mainly had been taken into consideration to assess pressure vessel service life limitations, but when extending the service life up to 60 years and more the non-hardening mechanism (intergranular embrittlement of the steels) must be taken into account as well. In this connection NRC “Kurchatov Institute” has initiated a number of works on investigations of this mechanism contribution to the total embrittlement of reactor pressure vessel steels. The main results of these investigations are described in this article. Results of grain boundary phosphorus concentration measurements in specimens made of first generation of VVER-type pressure vessels materials as well as VVER-1000 surveillance specimens are presented. An assessment of non-hardening mechanism contribution to the total ductile-to- brittle transition temperature shift is given.

  3. Optimizing the Design of Small Fast Spectrum Battery-Type Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Staffan Qvist

    2014-07-01

    Full Text Available This study is focused on defining and optimizing the design parameters of inherently safe “battery” type sodium-cooled metallic-fueled nuclear reactor cores that operate on a single stationary fuel loading at full power for 30 years. A total of 29 core designs were developed with varying power and flow conditions, including detailed thermal-hydraulic, structural-mechanical and neutronic analysis. Given set constraints for irradiation damage, primary cycle pressure drop and inherent safety considerations, the attainable power range and performance characteristics of the systems are defined. The optimum power level for a core with a coolant pressure drop limit of 100 kPa and an irradiation damage limit of 200 DPA (displacements per atom is found to be 100 MWt/40 MWe. Raising the power level of an optimized core gives significantly higher attainable power densities and burnup, but severely decreases safety margins and increases the irradiation damage. A fully optimized inherently safe battery-type fast reactor core with an active height and diameter of 150 cm (2.6 m3, a pressure drop limit of 100 kPa and an irradiation damage limit of 300 DPA can be designed to operate at 150 MWt/60 MWe for 30 years, reaching an average discharge burnup of 100 MWd/kg-actinide.

  4. Design and manufacture of a D-shape coil-based toroid-type HTS DC reactor using 2nd generation HTS wire

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kwangmin, E-mail: kwangmin81@gmail.com [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Go, Byeong-Soo; Sung, Hae-Jin; Park, Hea-chul; Kim, Seokho [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of); Jin, Yoon-Su; Oh, Yunsang [Vector Fields Korea Inc., Pohang 790-834 (Korea, Republic of); Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of)

    2014-09-15

    Highlights: • The authors designed and fabricated a D-shape coil based toroid-type HTS DC reactor using 2G GdBCO HTS wires. • The toroid-type magnet consisted of 30 D-shape double pancake coil (DDC)s. The total length of the wire was 2.32 km. • The conduction cooling method was adopted for reactor magnet cooling. • The maximum cooling temperature of reactor magnet is 5.5 K. • The inductance was 408 mH in the steady-state condition (300 A operating). - Abstract: This paper describes the design specifications and performance of a real toroid-type high temperature superconducting (HTS) DC reactor. The HTS DC reactor was designed using 2G HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The target inductance of the HTS DC reactor was 400 mH. The expected operating temperature was under 20 K. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. Performances of the toroid-type HTS DC reactor were analyzed through experiments conducted under the steady-state and charge conditions. The fundamental design specifications and the data obtained from this research will be applied to the design of a commercial-type HTS DC reactor.

  5. Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

    Energy Technology Data Exchange (ETDEWEB)

    1994-03-01

    On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci{sup 192}Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape.

  6. Thermal-hydraulic Fortran program for steady-state calculations of plate-type fuel research reactors

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

    Full Text Available The safety assessment of research and power reactors is a continuous process covering their lifespan and requiring verified and validated codes. Power reactor codes all over the world are well established and qualified against real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume a lot of running time. On the other hand, most research reactor codes still require much more data for validation and qualification. It is, therefore, of benefit to any regulatory body to develop its own codes for the review and assessment of research reactors. The present paper introduces a simple, one-dimensional Fortran program called THDSN for steady-state thermal-hydraulic calculations of plate-type fuel research reactors. Besides calculating the fuel and coolant temperature distributions and pressure gradients in an average and hot channel, the program calculates the safety limits and margins against the critical phenomena encountered in research reactors, such as the onset of nucleate boiling, critical heat flux and flow instability. Well known thermal-hydraulic correlations for calculating the safety parameters and several formulas for the heat transfer coefficient have been used. The THDSN program was verified by comparing its results for 2 and 10 MW benchmark reactors with those published in IAEA publications and a good agreement was found. Also, the results of the program are compared with those published for other programs, such as the PARET and TERMIC.

  7. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  8. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  9. Propagation of cracks by stress corrosion in conditions of BWR type reactor; Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua en ebullicion (BWR)

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, F.J. [ININ, 52045 Estado de Mexico (Mexico); Fuentes C, P. [ITT, Metepec, Estado de Mexico (Mexico)]. E-mail: fjmc@nuclear.inin.mx

    2004-07-01

    In this work, the obtained results when applying the Hydrogen Chemistry to a test tube type Compact Tension (CT), built in austenitic stainless steel 304l, simulating the conditions to those that it operates a Boiling Water Reactor (BWR), temperature 288 C and pressure of 8 MPa are presented. With the application of this water chemistry, seeks to be proven the diminution of the crack propagation speed. (Author)

  10. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  11. 池式调谐质量阻尼器的动力分析%Dynamic Analysis on Pool-type Tuned Mass Damper

    Institute of Scientific and Technical Information of China (English)

    周福霖; 黄东阳; 谭平

    2007-01-01

    Presented a pool-type tuned mass damper (TMD), in which the synthetical effects of both tuned liquid damper (TLD) and TMD were considered simultaneously. The equations of motion of the structure with a pool-type TMD, TMD or TLD were formulated. An indoor swimming pool was taken as an example to investigate the control performance of this new pool-type TMD. The control effectivenesses of the pool-type TMD system with various pool-lengths along vibration direction were also computed and compared with that of the corresponding TMD, TLD system under wind or earthquake excitations respectively. The suggestions for designing a pooltype TMD were provided through phase difference analyses. The results show that the optimally designed pool-type TMD can achieve a performance level which is close to that of an ordinary TMD, while both TMDs far outperform TLD. Compared with ordinary TMD, the pool-type TMD has more economical and practical values.%定义了池式调谐质量阻尼器(TMD),运用调谐液体阻尼器(TLD)与TMD相结合的方法综合考虑其总体动力效应;同时推导了结构与TMD、TLD以及池式TMD系统的运动方程.以结构物室内游泳池为例,研究了不同池长设计值的池式TMD在风振与地震激励下对结构的控制性能,对比了相应TMD、TLD的控制效果,并运用相位差分析的方法对池式TMD的设置提出了建议.结果表明:池式TMD经优化设计后可取得与普通TMD相近的控制效果,均远优于TLD;同时,池式TMD较普通TMD更经济且更具实用价值.

  12. Aging of reactor vessels in LWR type reactors; Envejecimiento de la vasija y de los internos del nuclear de los reactores tipo LWR

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.

    2004-07-01

    Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs.

  13. Achievement of treatment goals with canagliflozin in patients with type 2 diabetes mellitus: a pooled analysis of randomized controlled trials.

    Science.gov (United States)

    Blonde, Lawrence; Woo, Vincent; Mathieu, Chantal; Yee, Jacqueline; Vijapurkar, Ujjwala; Canovatchel, William; Meininger, Gary

    2015-11-01

    To evaluate attainment of diabetes-related treatment goals with canagliflozin, a sodium glucose co-transporter 2 inhibitor, versus placebo in patients with type 2 diabetes mellitus (T2DM). Data were pooled from four 26-week, placebo-controlled, Phase 3 studies of patients with T2DM (N = 2313). Goal attainment with canagliflozin 100 and 300 mg versus placebo was evaluated in the overall population, and in subgroups based on age and sex, at baseline and Week 26. ClinicalTrials.gov NCT01081834, NCT01106677, NCT01106625, NCT01106690. Proportion of patients achieving hemoglobin A1C (A1C) canagliflozin 100 and 300 mg compared with placebo. More patients achieved body weight reduction of ≥ 5% with canagliflozin 100 and 300 mg versus placebo at Week 26. Fewer patients had LDL-C canagliflozin 100 and 300 mg versus placebo. Canagliflozin 100 and 300 mg also provided better attainment of the composite endpoint of A1C Canagliflozin was associated with better attainment of diabetes-related treatment goals compared with placebo, and was generally well tolerated at 26 weeks.

  14. Ma-Pi 2 macrobiotic diet and type 2 diabetes mellitus: pooled analysis of short-term intervention studies.

    Science.gov (United States)

    Porrata-Maury, C; Hernández-Triana, M; Ruiz-Álvarez, V; Díaz-Sánchez, M E; Fallucca, F; Bin, W; Baba-Abubakari, B; Pianesi, M

    2014-03-01

    The macrobiotic, Ma-Pi 2 diet (12% protein, 18% fat and 70% carbohydrate), has shown benefit in adults with type 2 diabetes mellitus (T2DM). This pooled analysis aims to confirm results from four, 21-day intervention studies with the Ma-Pi 2 diet, carried out in Cuba, China, Ghana and Italy. Baseline and end of study biochemical, body composition and blood pressure data, were compared using multivariate statistical methods and assessment of the Cohen effect size (d). Results showed that all measured indicators demonstrated significant changes (p diet was Italy (1.96), China (1.79), Cuba (1.38) and Ghana (0.98). The magnitude of the individual effect on each variable by country, and the global effect by country, was independent of the sample size (p > 0.05). Similarly, glycemia and glycemic profiles in all four studies were independent of the sample size (p = 0.237). The Ma-Pi diet 2 significantly reduced glycemia, serum lipids, uremia and cardiovascular risk in adults with T2DM. These results suggest that the Ma-Pi 2 diet could be a valid alternative treatment for patients with T2DM and point to the need for further clinical studies. Mechanisms related to its benefits as a functional diet are discussed.

  15. [Continuous operation of hydrogen bio-production reactor with ethanol-type fermentation].

    Science.gov (United States)

    Ren, Nan-qi; Gong, Man-li; Xing, De-feng

    2004-11-01

    The natural response of a continuous stirred tank reactor (CSTR) for hydrogen bio-production using molasses wastewater as substrate was investigated. Emphasis was placed on assessing the operational controlling strategy on the stable operation of CSTR with high efficiency. It was found that at an initial biomass of 15g/L, an equilibrial microbial community in the ethanol-type fermentation and efficient stable operation of CSTR could be established with following conditions: temperature of 35 degrees C +/- 1 degrees C, COD organic loading rate (OLR) of 40kg/(m3 x d), hydraulic retention time (HRT) of 4h, pH value of 4.6 - 4.9 and oxidation reduction potential (ORP) of -450 - -470mV. Following that, hydrogen production in the reactor was relatively stable. The observed maximal hydrogen bio-production rate was 7.63m3/(m3 x d). The content of hydrogen in the biogas was about 40% - 58%. COD removal rate was between 22% - 26%. The total content of ethanol and acetic acid in the fermentative end products was above 80%.

  16. Proposal for a novel type of small scale aneutronic fusion reactor

    Science.gov (United States)

    Gruenwald, J.

    2017-02-01

    The aim of this work is to propose a novel scheme for a small scale aneutronic fusion reactor. This new reactor type makes use of the advantages of combining laser driven plasma acceleration and electrostatic confinement fusion. An intense laser beam is used to create a lithium-proton plasma with high density, which is then collimated and focused into the centre of the fusion reaction chamber. The basic concept presented here is based on the 7Li-proton fusion reaction. However, the physical and technological fundamentals may generally as well be applied to 11B-proton fusion. The former fusion reaction path offers higher energy yields while the latter has larger fusion cross sections. Within this paper a technological realisation of such a fusion device, which allows a steady state operation with highly energetic, well collimated ion beam, is presented. It will be demonstrated that the energetic break even can be reached with this device by using a combination of already existing technologies.

  17. Efficacy and Safety of Canagliflozin in Individuals Aged 75 and Older with Type 2 Diabetes Mellitus: A Pooled Analysis.

    Science.gov (United States)

    Sinclair, Alan J; Bode, Bruce; Harris, Stewart; Vijapurkar, Ujjwala; Shaw, Wayne; Desai, Mehul; Meininger, Gary

    2016-03-01

    To compare the efficacy and safety of canagliflozin, a sodium glucose co-transporter 2 inhibitor developed to treat type 2 diabetes mellitus (T2DM), in individuals younger than 75 and those aged 75 and older. Randomized Phase 3 studies. International study centers. Adults with T2DM. Changes from baseline in glycosylated hemoglobin (HbA1c ), fasting plasma glucose (FPG), blood pressure (BP), and body weight were measured. Efficacy was evaluated using pooled data from six randomized, double-blind, placebo-controlled studies (N = 4,158; n = 3,975 aged Canagliflozin 100 and 300 mg were associated with placebo-subtracted mean reductions in HbA1c in participants younger than 75 (-0.69% and -0.85%, respectively) and aged 75 and older (-0.65% and -0.55%, respectively). Dose-related reductions in FPG, body weight, and BP were seen with canagliflozin 100 and 300 mg in participants in both age groups. Overall AE incidence was 67.1% with canagliflozin 100 mg, 68.6% with canagliflozin 300 mg, and 65.9% with non-canagliflozin (pooled group of comparators in all studies) in participants younger than 75, and 72.4%, 79.1%, and 72.3%, respectively, in those aged 75 and older, with a similar safety profile in both groups. The incidence of volume depletion-related AEs was 2.2%, 3.1%, and 1.4% in participants younger than 75 with canagliflozin 100 and 300 mg and non-canagliflozin, respectively, and 4.9%, 8.7%, and 2.6%, respectively, in those aged 75 and older. Canagliflozin improved glycemic control, body weight, and BP in participants aged 75 and older. The overall incidence of AEs was high across treatment groups in participants aged 75 and older and higher than in those younger than 75. The safety profile of canagliflozin was generally similar in both age groups, with a higher incidence of AEs related to volume depletion observed with canagliflozin in participants aged 75 and older than in those younger than 75. These findings support canagliflozin, starting with the 100

  18. The application of research reactor Maria for analysis of thorium use in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Chwaszczewski, S.; Andrzejewski, K.; Myslek-Laurikainen, B.; Pytel, B.; Szczurek, J. [Dep. Thorium Project, Institute of Atomic Energy POLATOM, 05-400 Otwock-Swierk (Poland); Polkowska-Motrenko, H. [Institute of Nuclear Chemistry and Technology, ul.Dorodna 16 03-195 Warszawa (Poland)

    2010-07-01

    The MARIA reactor, pool-type light-water cooled and beryllium moderated nuclear research reactor was used to evaluate the {sup 233}U breeding during the experimental irradiation of the thorium samples. The level of impurities concentrations was determined using ICP-MS method. The associated development of computer programs for analysis of application of thorium in EPR reactor consist of PC version of CORD-2/GNOMER system are presented. (authors)

  19. You can pool faecal samples from individual pigs to test for Porcine Circovirus Type 2 and Lawsonia intracellularis using real-time PCRs

    DEFF Research Database (Denmark)

    Holyoake, Patricia K.; Hjulsager, Charlotte Kristiane; Larsen, Lars Erik;

    Introduction Real-time PCR tests have been developed to detect and quantify Porcine Circovirus type 2 (PCV2) and Lawsonia intracellularis in pigs’ faeces. Pooling of individual faecal samples is often used to reduce the costs of diagnostic testing. The objective of this study was to evaluate any ...

  20. Experimental investigation of heat transfer during severe accident of a Pressurized Heavy Water Reactor with simulated decay heat generation in molten pool inside calandria vessel

    Energy Technology Data Exchange (ETDEWEB)

    Prasad, Sumit Vishnu, E-mail: svprasad@barc.gov.in; Nayak, Arun Kumar, E-mail: arunths@barc.gov.in

    2016-07-15

    Highlights: • Scaled test facility simulating the calandria vessel and calandria vault water of PHWR with simulated decay heat was built. • Experiments conducted with simulant material at about 1200 °C. • Experimental result shows that melt coolability and growth rate of crust thickness are affected by presence of decay heat. • No gap was observed between the crust and vessel on opening. • Result shows that vessel integrity is intact with presence of water inside water tank in both cases. - Abstract: The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat in the simulated calandria vessel. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1200 °C. Decay heat in the melt pool was simulated using four high watt heaters cartridges, each having 9.2 kW. The temperature distributions inside the molten pool, across the vessel wall thickness and vault water were measured. Experimental results obtained are compared with the results obtained previously for no decay heat case. The results indicated that presence of decay heat seriously affects the coolability behaviour and formation of crust in the melt pool. The location and magnitude of maximum heat flux and surface temperature of the vessel also are affected in the presence of decay heat.

  1. Synthesis of layered double hydroxide nanosheets by coprecipitation using a T-type microchannel reactor

    Science.gov (United States)

    Pang, Xiujiang; Sun, Meiyu; Ma, Xiuming; Hou, Wanguo

    2014-02-01

    The synthesis of Mg2Al-NO3 layered double hydroxide (LDH) nanosheets by coprecipitation using a T-type microchannel reactor is reported. Aqueous LDH nanosheet dispersions were obtained. The LDH nanosheets were characterized by X-ray diffraction, transmission electron microscopy, atomic force microscopy and particle size analysis, and the transmittance and viscosity of LDH nanosheet dispersions were examined. The two-dimensional LDH nanosheets consisted of 1-2 brucite-like layers and were stable for ca. 16 h at room temperature. In addition, the co-assembly between LDH nanosheets and dodecyl sulfate (DS) anions was carried out, and a DS intercalated LDH nanohybrid was obtained. To the best of our knowledge, this is the first report of LDH nanosheets being directly prepared in bulk aqueous solution. This simple, cheap method can provide naked LDH nanosheets in high quantities, which can be used as building blocks for functional materials.

  2. Large pool LMFBR design. Final report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Wett, J. F.; Churchill, J. R.

    1979-03-01

    The design effort reported is an extension on past design effort and continuous concentration on those parts of the nuclear island unique to a commercial size pool type LMFBR. In particular, the work covers the reactor vessel, deck, rotating plugs, upper and lower internals, internal plenum separator system, IHX, pumps, cold traps, intermediate system layout, containment/confinement system, plot plan, and residual heat removal systems. Preliminary thermal, hydraulic, stress, and system analyses are also presented.

  3. Influence of irradiation by {sup 60}Co gamma-quanta on reliability of IR-LEDs based upon AlGaAs heterostructures

    Energy Technology Data Exchange (ETDEWEB)

    Gradoboev, Alexander V.; Simonova, Anastasiia V.; Orlova, Ksenia N. [National Research Tomsk Polytechnic University, Lenina avenue 30, 634050 Tomsk (Russian Federation)

    2016-12-15

    We consider the influence of preliminary irradiation by {sup 60}Co gamma quanta on emission power decrease during the operating time of IR-LEDs based upon AlGaAs double heterostructures. Irradiation was realized in passive power mode of the LEDs prior to aging. Aging under long operating time conditions was simulated by a step-stress approach. We have determined that the emissive power decrease of LED during operating time has two stages. On the first stage, decrease of LED emissive power is due to rearrangement of original defect structure. On the second stage, emissive power goes down as the result of inducing new structural defects. We have shown identical multistage mechanisms of emissive power drop are observed during both operating time and influence of ionizing radiation. We have established that preliminary irradiation has increased reliability and operating time of LEDs. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  4. Vernal Pools

    Data.gov (United States)

    California Department of Resources — This is a polygon layer representing existing vernal pool complexes in California's Central Valley, as identified and mapped by Dr. Robert F. Holland. The purpose of...

  5. Radiotoxicity and decay heat power of spent nuclear fuel of VVER type reactors at long-term storage.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Radiotoxicity and decay heat power of the spent nuclear fuel of VVER-1000 type reactors are calculated during storage time up to 300,000 y. Decay heat power of radioactive waste (radwaste) determines parameters of the heat removal system for the safe storage of spent nuclear fuel. Radiotoxicity determines the radiological hazard of radwaste after its leakage and penetration into the environment.

  6. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  7. Use of plutonium in PWR-type reactors; Utilisation du plutonium dans les REP

    Energy Technology Data Exchange (ETDEWEB)

    Berthet, A. [Electricite de France (EDF), 75 - Paris (France). Direction de l' Equipement

    1999-04-01

    The plutonium is used, as fuel, in the pressurized water reactors. It does not exist in nature; butit is fabricated in the reactor by neutrons capture. The MOX (Mixed Oxides) is its usual name. A part is consumed by the fission, the remainder is found in the used fuel released from the reactor. The paper deals with the plutonium specificities, the research and development programs about this fuel. The technical specifications of the PWR recycling the plutonium are also included (radiation protection, reactor fueling). (A.L.B.)

  8. Formation of a nuclear reactor's molten core bath in a crucible-type corium catcher for a nuclear power station equipped with VVER reactors

    Science.gov (United States)

    Beshta, S. V.; Vitol', S. A.; Granovskii, V. S.; Kalyago, E. K.; Kovtunova, S. V.; Krushinov, E. V.; Sulatskaya, M. B.; Sulatskii, A. A.; Khabenskii, V. B.; Al'Myashev, V. I.; Gusarov, V. V.

    2011-05-01

    Results from a calculation study on analyzing the formation of a melt bath in a crucible-type catcher for the conditions of a severe accident at a nuclear power station equipped with VVER-1000 reactors are presented. It is shown that the heat loads exerted on the water-cooled walls of the corium catcher shell are limited to a permissible level at which the necessary margins to nucleate boiling crisis and to destruction are ensured under the conditions of thermal and mechanical loading of the shell. An important role of sacrificial material in the efficient operation of the corium catcher is pointed out.

  9. Experimental evaluation of two different types of reactors for CO2 removal from gaseous stream by bottom ash accelerated carbonation.

    Science.gov (United States)

    Lombardi, L; Carnevale, E A; Pecorini, I

    2016-12-01

    Low methane content landfill gas may be enriched by removing carbon dioxide. An innovative process, based on carbon dioxide capture and storage by means of accelerated carbonation of bottom ash is proposed and studied for the above purpose. Within this research framework we devoted a preliminary research activity to investigate the possibility of improving the way the contact between bottom ash and landfill gas takes place: this is the scope of the work reported in this paper. Two different types of reactors - fixed bed and rotating drum - were designed and constructed for this purpose. The process was investigated at laboratory scale. As the aim of this phase was the comparison of the performances of the two different reactors, we used a pure stream of CO2 to preliminarily evaluate the reactor behaviors in the most favorable condition for the process (i.e. maximum CO2 partial pressure at ambient condition). With respect to the simple fixed bed reactor concept, some modifications were proposed, consisting of separating the ash bed in three layers. With the three layer configuration we would like to reduce the possibility for the gas to follow preferential paths through the ash bed. However, the results showed that the process performances are not significantly influenced by the multiple layer arrangement. As an alternative to the fixed bed reactor, the rotating drum concept was selected in order to provide continuous mixing of the solids. Two operating parameters were considered and varied during the tests: the filling ratio and the rotating speed. Better performances were observed for lower filling ratio while the rotating speed showed minor importance. Finally the performances of the two reactors were compared. The rotating drum reactor is able to provide improved carbon dioxide removal with respect to the fixed bed one, especially when the rotating reactor is operated at low filling ratio values and slow rotating speed values. Comparing the carbon dioxide

  10. Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

    CSIR Research Space (South Africa)

    Hirschberg, G

    1999-03-01

    Full Text Available of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg G abor Hirschberg a,P al Baradlai a,K alm an Varga a,*, Gerrit Myburg b, J anos Schunk c,P eter Tilky c, Paul Stoddart d a Department of Radiochemistry, University...-cooled nuclear reactors is of great importance for a number of practical reasons. For instance, under normal operating conditions (when there is no ?ssion product release due to fuel cladding failure) the majority of radioactive contamination in the pri- mary...

  11. Steam feed and effect of steam-thermal seal in thermolysis of tire shreds in a screw-type reactor

    Science.gov (United States)

    Kalitko, V. A.

    2010-05-01

    On the basis of experience in commercial operation, the effect of steam seal in tire-shred pyrolysis in a screw-type reactor with superheated steam has been considered and analytically substantiated; there, local steam feed produces the above effect at the total reduced pressure and keeps air from entering the reactor without sluices or valves used for hermetization of its loading and unloading. It has been shown that the increase in the production rate of pyrolysis due to the heating by steam amounts to 10-15% and is limited by the diffusion transfer in the reactor’s charge bed.

  12. Major types of dietary fat and risk of coronary heart disease: a pooled analysis of 11 cohort studies

    DEFF Research Database (Denmark)

    Jakobsen, Marianne Uhre; O'Reilly, Eilis J; Heitmann, Berit Lilienthal

    2009-01-01

    fatty acids or carbohydrates should replace energy from SFAs to prevent CHD. DESIGN: This was a follow-up study in which data from 11 American and European cohort studies were pooled. The outcome measure was incident CHD. RESULTS: During 4-10 y of follow-up, 5249 coronary events and 2155 coronary deaths...

  13. GGPPS-mediated Rab27A geranylgeranylation regulates β cell dysfunction during type 2 diabetes development by affecting insulin granule docked pool formation.

    Science.gov (United States)

    Jiang, Shan; Shen, Di; Jia, Wen-Jun; Han, Xiao; Shen, Ning; Tao, Weiwei; Gao, Xiang; Xue, Bin; Li, Chao-Jun

    2016-01-01

    Loss of first-phase insulin secretion associated with β cell dysfunction is an independent predictor of type 2 diabetes mellitus (T2DM) onset. Here we found that a critical enzyme involved in protein prenylation, geranylgeranyl pyrophosphate synthase (GGPPS), is required to maintain first-phase insulin secretion. GGPPS shows a biphasic expression pattern in islets of db/db mice during the progression of T2DM: GGPPS is increased during the insulin compensatory period, followed by a decrease during β cell dysfunction. Ggpps deletion in β cells results in typical T2DM β cell dysfunction, with blunted glucose-stimulated insulin secretion and consequent insulin secretion insufficiency. However, the number and size of islets and insulin biosynthesis are unaltered. Transmission electron microscopy shows a reduced number of insulin granules adjacent to the cellular membrane, suggesting a defect in docked granule pool formation, while the reserve pool is unaffected. Ggpps ablation depletes GGPP and impairs Rab27A geranylgeranylation, which is responsible for the docked pool deficiency in Ggpps-null mice. Moreover, GGPPS re-expression or GGPP administration restore glucose-stimulated insulin secretion in Ggpps-null islets. These results suggest that GGPPS-controlled protein geranylgeranylation, which regulates formation of the insulin granule docked pool, is critical for β cell function and insulin release during the development of T2DM.

  14. Implications of reactor type and conditions on first-order hydrolysis rate assessment of maize silage.

    Science.gov (United States)

    Pabón Pereira, C P; Zeeman, G; Zhao, J; Ekmekci, B; van Lier, J B

    2009-01-01

    The biodegradability and first-order hydrolysis coefficient of maize silage have been assessed from batch experiments using different types of inoculum and substrate to inocula (S/I) ratios, and from CSTRs working at different hydraulic retention times (HRTs). In the batch experiments, the assessed maximum biodegradability of the maize silage was 68 (+/-2.7)% and 73(+/-2.9)% while the first order hydrolysis was 0.26 (+/-0.01) and 0.27(+/-0.02) d(-1), using granular and a mixture of granular and suspended inoculum, respectively. In the CSTR experiment biodegradability ranged from 41-65% depending on the HRT applied whereas the calculated first order hydrolysis coefficient was 0.32 d(-1). It is concluded that batch experiments can be used to assess first order hydrolysis constants and biodegradability provided that a well balanced inoculum is guaranteed. Further, it is shown that CSTR reactors digesting maize silage and operating at HRTs as low as 20 days can attain 88% of maximum biodegradability as long as pH fluctuations are minimized. 2 mmol NaHCO3 per gram maize silage was calculated to suffice for the purpose.

  15. Intercomparison of Different Types of Locally Prepared Concretes and Its Usability for Reactor Neutron Shielding

    Science.gov (United States)

    El-Kolaly, M. A.; Makarious, A. S.; Bashter, I. I.; Kansouh, W. A.

    Measurements have been carried out to study the attenuation of neutron from a horizontal channel of the ET-RR-1 reactor. The assessments of neutron distribution inside three different types of locally prepared concretes have been evaluated.Neutron intensities in ilmenite-limonite concrete shield show an exponential decrease with increasing concrete thickness. Ilmenite concrete is a good attenuator for thermal and intermediate neutrons. However, ordinary and ilmenite-limonite concretes show efficient shielding for fast neutrons.Translated AbstractVergleich verschiedener Zementarten hinsichtlich ihrer Brauchbarkeit zur Neutronenabschirmung von ReaktorenMessungen zur Untersuchung der Neutronenabschwächung in einem horizontalen Kanal eines ET-RR-1-Reaktors wurden durchgeführt. Die Charakteristika der Neutronenverteilung innerhalb dreier unterschiedlich zusammengesetzter Zemente wurden bestimmt. Die Neutronenintensität in einem Schild aus Ilmenite-Limonitezement zeigt einen exponentiellen Abfall mit wachsender Dicke. Ilmenitezement ist ein guter Schild für thermale und mittlere Neutronen. Normaler und Ilmenite-Limonitezement zeigen effektive Abschirmung bei schnellen Neutronen.

  16. Espacio Pool

    OpenAIRE

    2013-01-01

    Espacio Pool es un grupo abierto de usuarios vinculados a la Facultad de Bellas Artes UCM que actúa sobre los márgenes que la circundan. Toma su nombre de los estanques vacíos ubicados en los jardines de la cafetería, y elige el término anglosajón por sus connotaciones relativas a lo participativo y autogestionado. Su objetivo es habitar los espacios de “vacuidad” para que sean efectivos y permutadores. Espacio Pool organiza la celebración de encuentros a partir de la construcción de disposit...

  17. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

  18. Effect of conditions of air-lift type reactor work on cadmium adsorption

    Energy Technology Data Exchange (ETDEWEB)

    Filipkowska, Urszula; Szymczyk, Paula Szymczyk; Kuczajowska-Zadrozna, Malgorzata; Joezwiak, Tomasz [University of Warmia and Mazury in Olsztyn, Warszawska (Poland)

    2015-10-15

    We investigated cadmium sorption by activated sludge immobilized in 1.5% sodium alginate with 0.5% polyvinyl alcohol. Experiments were conducted in an air-lift type reactor at the constant concentration of biosorbent reaching 5 d.m./dm{sup 3}, at three flow rates: 0.1, 0.25 and 0.5 V/h, and at three concentrations of the inflowing cadmium solution: 10, 25 and 50mg/dm{sup 3}. Analyses determined adsorption capacity of activated sludge immobilized in alginate as well as reactor's work time depending on flow rate and initial concentration of the solution. Results achieved were described with the use of Thomas model. The highest adsorption capacity of the sorbent (determined from the Thomas model), i.e., 200.2mg/g d.m. was obtained at inflowing solution concentration of 50mg/dm{sup 3} and flow rate of 0.1V/h, whereas the lowest one reached 53.69mg/g d.m. at the respective values of 10mg/dm{sup 3} and 0.1 V/h. Analyses were also carried out to determine the degree of biosorbent adsorption capacity utilization at the assumed effectiveness of cadmium removal - at the breakthrough point (C=0.05*C{sub 0}) and at adsorption capacity depletion point (C−0.9*C0). The study demonstrated that the effectiveness of adsorption capacity utilization was influenced by both the concentration and flow rate of the inflowing solution. The highest degree of sorbent capacity utilization was noted at inflowing solution concentration of 50mg/dm{sup 3} and flow rate of 0.1 V/h, whereas the lowest one at the respective values of 10mg/dm{sup 3} and 0.1 V/h. The course of the process under dynamic conditions was evaluated using coefficients of tangent inclination - a, at point C/C{sub 0}=1/2. A distinct tendency was demonstrated in changes of tangent slope a as affected by the initial concentration of cadmium and flow rate of the solution. The highest values of a coefficient were achieved at the flow rate of 0.1 V/h and initial cadmium concentration of 50mg/dm{sup 3}.

  19. Technology of Anodization of Aluminum Pool Shell for Nuclear Reactor in Oxalic Acid Solution%草酸溶液中铝池壳的阳极氧化技术

    Institute of Scientific and Technical Information of China (English)

    白新德; 白光美; 郭金梁; 陈鹤鸣; 马春来; 彭德全; 董铎; 钟大辛; 陆金法; 郭宝华; 周昕

    2004-01-01

    A lot of experiments about the anodization of A0 pure aluminum in oxalic acid solution were carried out. The technology parameters were decided, including the anode current density, the oxidization time and the temperature of the electrolyte solution. During the anode oxidization of the pool shell, some special key technologies were solved, including the oxidization technology of large equipment, the technology of the treatment with layer by layer and sealing with thin plastic films, the selection of the power, the circular cooling of the electrolyte solution, etc. Treating the whole pool shell with such measures, the abilities of the corrosion-proof and protection were greatly improved. The nuclear reactor has run for 37 years (from 1964 to 2001), by now the oxidization film on the surface of the pool shell is still bright. It can confirm that treating of the whole pool shell is successful and necessary, which can provide the help and reference for building the same model nuclear reactors.%清华大学核能研究院屏蔽实验反应堆池壳用A0纯铝制成,并于1964年对池壳进行了全部阳极氧化处理.至2001年,反应堆运行了37年,池壳表面仍然有一层发亮的氧化膜,与国内外同类型的未经阳极氧化处理的反应堆相比较,说明氧化膜极好地保护了池壳.该反应堆池壳容积大于50 m3.表面积达100 m2,把这样大的铝制设备在完全安装完毕后进行全部表面阳极氧化处理,在处理工艺上是很困难的.本文阐明了解决这些问题的方法.对A0纯铝在草酸中的阳极氧化做了许多实验,获得了A0纯铝在3%(w/%)草酸中的阳极氧化的基本规律,并确定了最终的工艺参数.包括氧化电流密度,氧化时间和电解液的温度.在铝池壳的阳极氧化过程中解决了许多关键的工艺:如大型工件阳极氧化工艺、氧化机制、性能;分层处理与薄膜密封技术;电源的选择;电解液的循环冷却;氧化膜的质量检测等.

  20. Large-scale surface dielectric barrier discharge type reactor : effect of the electric wind on the conversion effectiveness

    Energy Technology Data Exchange (ETDEWEB)

    Jolibois, J. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique; Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Zouzou, N.; Moreau, E. [Poitiers Univ., Futuroscope Chasseneuil Cedex (France). Centre national de la recherche scientifique, Inst. Pprime; Tatibouet, J.M. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique

    2010-07-01

    Non-thermal plasma (NTP) techniques offer an innovative approach for air pollution reduction. Most studies in NTP techniques use volumetric discharge reactors with small dimensions and low flow rates at laboratory scale. The objective of this study was to develop an air pollution control plasma reactor at industrial scale with surface discharge. Propene (C{sub 3}H{sub 6}) was oxidized at high flow rates in a large-scale plasma reactor based on surface dielectric barrier discharge (DBD). Three different configurations of surface discharges were tested with 15 ppm of C{sub 3}H{sub 6} in air at ambient temperature for a flow rate of 50 m{sup 3} per hour. The properties of these different surface discharges were analyzed using chemical measurements and 3 component particle image velocimetry (PIV) measurements. PIV measurements were used characterize the effect of the electric wind on the polluted gas airflow inside the reactor and to explain the differences of effectiveness of the three tested plasma generators. For the three plasma generators, a propene oxidation of up to 45 percent was obtained at one J per liter. The electric wind produced by the surface discharge resulted in the formation of vortices inside the plasma reactor. This electric wind can increase gas mixing inside the plasma reactor and therefore plays a key role in conversion efficiency. It was concluded that the electric wind produced by surface discharges enables the use of this type of discharge for VOC elimination at high flow rate, with the same effectiveness of volumetric discharges. 5 refs., 10 figs.

  1. Stress corrosion (Astm G30-90 standard) in 08x18H10T stainless steel of nuclear fuel storage pool in WWER reactors; Corrosion bajo esfuerzo (Norma ASTM G30-90) en acero inoxidable 08x18H10T de piscinas de almacenamiento de combustible nuclear en reactores V.V.E.R

    Energy Technology Data Exchange (ETDEWEB)

    Herrera, V.; Zamora R, L. [Centro de Estudios Aplicados al Desarrollo Nuclear (Cuba)

    1997-07-01

    At the water storage of the irradiated nuclear fuel has been an important factor in its management. The actual pools have its walls covered with inoxidable steel and heat exchangers to dissipate the residual heat from fuel. It is essential to control the water purity to eliminate those conditions which aid to the corrosion process in fuel and at related components. The steel used in this research was obtained from an austenitic inoxidizable steel standardized with titanium 08x18H10T (Type 321) similar to one of the two steel coatings used to cover walls and the pools floor. the test consisted in the specimen deformation through an U ply according to the Astm G30-90 standard. The exposition of the deformed specimen it was realized in simulated conditions to the chemical regime used in pools. (Author)

  2. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    Science.gov (United States)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-11-01

    High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  3. The upgrade and conversion of the ET-RR-1 research reactor using plate type fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Ashoub, N. [Reactor Physics Dept., Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Saleh, H.G. [Faculty of Girls for Arts and Education, Ain-Shams Univ., Cairo (Egypt)

    2001-11-01

    The ET-RR-1 research reactor has been operated at 2 MW since 1961 using EK-10 fuel elements with 10% enriched uranium. The reactor has been used for nuclear applied research and isotope production. In order to upgrade the reactor power to a reasonable limit facing up-to-date uses, core conversion by a new type of fuel element available is necessary. Two fuel elements in plate type are suggested in this study to be used in the ET-RR-1 reactor core rather than the utilized ones. The first element has a dimension of 8 x 8 x 50 cm and consists of 19.7% enriched uranium, which is typical for that utilized in the ET-RR-2 reactor, but with a different length. The other element is proposed with a dimension of 7 x 7 x 50 cm and has the same uranium enrichment. To accomplish safety requirements for these fuel elements, thermal-hydraulic evaluation has been carried out using the PARET code. To reach a core conversion of the ET-RR-1 reactor with the above two types of fuel elements, neutronic calculations have been performed using WIMSD4, DIXY2 and EREBUS codes. Some important nuclear parameters needed in the physical design of the reactor were calculated and included in this study. (orig.) [German] Der ET-RR-1 Forschungsreaktor wird seit 1961 unter Verwendung von EK-10 Brennelementen mit einer Leistung von 2 MW betrieben. Der Reaktor wird in der angewandten Forschung und zur Isotopenherstellung eingesetzt. Um die Reaktorleistung im Hinblick auf eine zeitgemaesse Nutzung der Anlage in einem vernuenftigen Mass zu erhoehen, ist eine Umwandlung des Kerns durch Verwendung neuartiger Brennelemente noetig. In der vorliegenden Untersuchung wird vorgeschlagen, anstelle der z. Z. verwendeten Elemente zwei neue, plattenfoermige Brennelemente zu verwenden. Das erste Element hat eine Groesse von 8 x 8 x 50 cm und besteht aus 19,7% angereichertem Uran, was den im ET-RR-2 Reaktor verwendeten Elementen entspricht, allerdings mit einer anderen Groesse. Das zweite Element hat die gleiche

  4. Irradiation rigs in material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rozenblum, F.; Gonnier, C.; Bignan, G. [CEA, Research Centers of Saclay and Cadarache (France)

    2011-07-01

    Osiris is a research reactor with a thermal power of 70 MW. It is a light-water reactor, open-core pool type, the principal aim of which is to carry out tests and irradiate structural materials and fuel elements of nuclear power plants under a high flux of neutrons, and to produce radioisotopes. Osiris operates around 200 days a year, in cycles of varying lengths from 3 to 4 weeks. A shutdown of about 10 days between two cycles allows reloading the core with fuel. Mainly 2 types of irradiation device are present: capsules for materials irradiation (CHOUCA and IRMA devices) and fuels irradiation loops (GRIFFONOS and ISABELLE). Although Osiris is still providing experiments of very good quality, it is facing obsolescence due to its ageing. Osiris is planned to be shut down during next decade. Consequently, it has been decided to launch the construction of the Jules Horowitz Reactor (JHR) in Cadarache. JHR is a water cooled reactor which provides the necessary flexibility and accessibility to manage several highly instrumented experiments, reproducing different reactor environments (water, gas or liquid metal loops), generating transient regimes (key for safety). The JHR facility includes the reactor building, including core, cooling system and the experimental bunkers connected to the core through pool wall penetrations and the auxiliary building, including pools and hot cells necessary for the experimental irradiation process. JHR core is optimised to produce high fast neutron flux to study structural material ageing and high thermal neutrons flux for fuel experiments. The conception of this first fleet of devices integrates the operational experience accumulated by the existing MTR and specifically the Osiris one

  5. Numerical Analysis of Magnetic Force of Dry-Type Air-Core Reactor

    Institute of Scientific and Technical Information of China (English)

    LIUZhi-gang; GENGYing-san; WANGJian-hua

    2004-01-01

    This paper presents a coupled magnetic-circuit method for computing the magnetic force of air-core reactor under short-time current. The current and the magnetic flux density are computed first and then the magnetic force is obtained. Thus, the dynamic stability performance of air-core reactor can be analyzed at the design stage to reduce experimental cost and shorten the lead-time of product development.

  6. Code Development of Radioactive Aerosol Scrubbing in Pool-Injection Zone

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Hyun Joung; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jang, Dong Soon [Chungnam National University, Daejeon (Korea, Republic of)

    2015-10-15

    The pool scrubbing models were reviewed and an aerosol scrubbing code has been prepared to calculate decontamination factor through the injection zone. The developed code has been verified using the experimental results and evaluated parametrically on the input variables. In injection zone, the initial steam condensation was most effective mechanism for the aerosol removal, and the steam fraction and pool temperature were highly affected on the decontamination factor by initial steam condensation. The aerosol scrubbing code will be updated to evaluate the decontamination factor at rise zone and finally whole pool scrubber phenomena. If a severe accident occurs in a nuclear power plant (NPP), the aerosol and gaseous fission products might be produced in the reactor vessel, and then released to the environment after the containment failure. FCVS (Filtered Containment Venting System) is one of the severe accident mitigation systems for retaining the containment integrity by discharging the high-temperature and high-pressure fission products to the environment after passing through the filtration system. In general, the FCVS is categorized into two types, wet and dry types. The scrubbing pool could play an important role in the wet type FCVS because a large amount of aerosol is captured in the water pool. The pool scrubbing phenomena have been modelled and embedded in several computer codes, such as SPARC (Suppression Pool Aerosol Removal Code), BUSCA (BUbble Scrubbing Algorithm) and SUPRA (Suppression Pool Retention Analysis). These codes aim at simulating the pool scrubbing process and estimating the decontamination factors (DFs) of the radioactive aerosol and iodine gas in the water pool, which is defined as the ratio of initial mass of the specific radioactive material to final massy after passing through the water pool. The pool scrubbing models were reviewed and an aerosol scrubbing code has been prepared to calculate decontamination factor through the injection

  7. Brazilian multipurpose reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-07-01

    The Brazilian Multipurpose Reactor (RMB) Project is an action of the Federal Government, through the Ministry of Science Technology and Innovation (MCTI) and has its execution under the responsibility of the Brazilian National Nuclear Energy Commission (CNEN). Within the CNEN, the project is coordinated by the Research and Development Directorate (DPD) and developed through research units of this board: Institute of Nuclear Energy Research (IPEN); Nuclear Engineering Institute (IEN); Centre for Development of Nuclear Technology (CDTN); Regional Center of Nuclear Sciences (CRCN-NE); and Institute of Radiation Protection and Dosimetry (IRD). The Navy Technological Center in Sao Paulo (CTMSP) and also the participation of other research centers, universities, laboratories and companies in the nuclear sector are important and strategic partnerships. The conceptual design and the safety analysis of the reactor and main facilities, related to nuclear and environmental licensing, are performed by technicians of the research units of DPD / CNEN. The basic design was contracted to engineering companies as INTERTHECNE from Brazil and INVAP from Argentine. The research units from DPD/CNEN are also responsible for the design verification on all engineering documents developed by the contracted companies. The construction and installation should be performed by specific national companies and international partnerships. The Nuclear Reactor RMB will be a open pool type reactor with maximum power of 30 MW and have the OPAL nuclear reactor of 20 MW, built in Australia and designed by INVAP, as reference. The RMB reactor core will have a 5x5 configuration, consisting of 23 elements fuels (EC) of U{sub 3}Si{sub 2} dispersion-type Al having a density of up to 3.5 gU/cm{sup 3} and enrichment of 19.75% by weight of {sup 23{sup 5}}U. Two positions will be available in the core for materials irradiation devices. The main objectives of the RMB Reactor and the other nuclear and radioactive

  8. Mechanical, chemical and radiological characterization of the graphite of the UNGG reactors type; Caracterisation mecanique, chimique et radiologique du graphite des reacteurs de la filiere UNGG

    Energy Technology Data Exchange (ETDEWEB)

    Bresard, I.; Bonal, J.P

    2000-07-01

    In the framework of UNGG reactors type dismantling procedures, the characterization of the graphite, used as moderator, has to be realized. This paper presents the mechanical, chemical and radiological characterizations, the properties measured and gives some results in the case of the Bugey 1 reactor. (A.L.B.)

  9. The OPAL reactor

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.; Irwin, T. [Australian Nuclear Science and Technology Organisation, Sydney (Australia); Ordonez, J.P. [INVAP SE, Bariloche (Argentina)

    2007-07-01

    The project to provide a replacement for Australia's HIFAR reactor began with governmental approval in September 1997 and reached its latest milestone with the achievement of the first full power operation of the OPAL reactor in November 2006. OPAL is a pool-type reactor with a thermal power of 20 MW and a fuel enrichment maximum of 20 per cent. This has been a successful project for both ANSTO (Australian Nuclear Science and Technology Organisation) and the contractor INVAP SE. This project was characterised by extensive interaction with the project's stake-holders during project definition and the use of a performance-based turnkey contract which gave the contractor the maximum opportunity to optimise the design to achieve performance and cost effectiveness. The contactor provided significant in-house resources as well as capacity to manage an international team of suppliers and sub-contractors. A key contributor to the project's successful outcomes has been the development and maintenance of an excellent working relationship between ANSTO and INVAP project teams. Commissioning was undertaken in accordance with the IAEA recommended stages. This paper presents the approaches used to define the project requirements, to choose the supplier and to deliver the project. The main results of hot commissioning are reviewed and the problems encountered examined. Operational experience since hot commissioning is also reviewed.

  10. 21 CFR 1250.89 - Swimming pools.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Swimming pools. 1250.89 Section 1250.89 Food and... SANITATION Sanitation Facilities and Conditions on Vessels § 1250.89 Swimming pools. (a) Fill and draw swimming pools shall not be installed or used. (b) Swimming pools of the recirculation type shall be...

  11. Pyrolysis of aseptic packages (tetrapak) in a laboratory screw type reactor and secondary thermal/catalytic tar decomposition.

    Science.gov (United States)

    Haydary, J; Susa, D; Dudáš, J

    2013-05-01

    Pyrolysis of aseptic packages (tetrapak cartons) in a laboratory apparatus using a flow screw type reactor and a secondary catalytic reactor for tar cracking was studied. The pyrolysis experiments were realized at temperatures ranging from 650 °C to 850 °C aimed at maximizing of the amount of the gas product and reducing its tar content. Distribution of tetrapak into the product yields at different conditions was obtained. The presence of H2, CO, CH4, CO2 and light hydrocarbons, HCx, in the gas product was observed. The Aluminum foil was easily separated from the solid product. The rest part of char was characterized by proximate and elemental analysis and calorimetric measurements. The total organic carbon in the tar product was estimated by elemental analysis of tars. Two types of catalysts (dolomite and red clay marked AFRC) were used for catalytic thermal tar decomposition. Three series of experiments (without catalyst in a secondary cracking reactor, with dolomite and with AFRC) at temperatures of 650, 700, 750, 800 and 850 °C were carried out. Both types of catalysts have significantly affected the content of tars and other components in pyrolytic gases. The effect of catalyst on the tetrapack distribution into the product yield on the composition of gas and on the total organic carbon in the tar product is presented in this work.

  12. Properties of bio-oil generated by a pyrolysis of forest cedar residuals with the movable Auger-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Shun; Ebitani, Kohki, E-mail: ebitani@jaist.ac.jp [School of Materials Science, Japan Advanced Institute of Science and Technology, 1-1 Asahidai, Nomi, Ishikawa 923-1292 (Japan); Miyazato, Akio [Nanotechnology Center, Japan Advanced Institute of Science and Technology, 1-1 Asahidai, Nomi, Ishikawa 923-1292 (Japan)

    2016-02-01

    Our research project has developed the new movable reactor for bio-oil production in 2013 on the basis of Auger-type system. This package would be a great impact due to the concept of local production for local consumption in the hilly and mountainous area in not only Japan but also in the world. Herein, we would like to report the properties of the bio-oil generated by the developing Auger-type movable reactor. The synthesized bio-oil possessed C: 46.2 wt%, H: 6.5 wt%, N: wt%, S: <0.1 wt%, O: 46.8 wt% and H{sub 2}O: 18.4 wt%, and served a good calorific value of 18.1 MJ/kg. The spectroscopic and mass analyses such as FT-IR, GC-MS, {sup 13}C-NMR and FT-ICR MS supported that the bio-oil was composed by the fine mixtures of methoxy phenols and variety of alcohol or carboxylic acid functional groups. Thus, it is suggested that the bio-oil generated by the new movable Auger-type reactor has a significant potential as well as the existing bio-oil reported previously.

  13. Related activities on management of ageing of Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham Van Lam [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

    1998-10-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)

  14. A modular method for the extraction of DNA and RNA, and the separation of DNA pools from diverse environmental sample types

    Science.gov (United States)

    Lever, Mark A.; Torti, Andrea; Eickenbusch, Philip; Michaud, Alexander B.; Šantl-Temkiv, Tina; Jørgensen, Bo Barker

    2015-01-01

    A method for the extraction of nucleic acids from a wide range of environmental samples was developed. This method consists of several modules, which can be individually modified to maximize yields in extractions of DNA and RNA or separations of DNA pools. Modules were designed based on elaborate tests, in which permutations of all nucleic acid extraction steps were compared. The final modular protocol is suitable for extractions from igneous rock, air, water, and sediments. Sediments range from high-biomass, organic rich coastal samples to samples from the most oligotrophic region of the world's oceans and the deepest borehole ever studied by scientific ocean drilling. Extraction yields of DNA and RNA are higher than with widely used commercial kits, indicating an advantage to optimizing extraction procedures to match specific sample characteristics. The ability to separate soluble extracellular DNA pools without cell lysis from intracellular and particle-complexed DNA pools may enable new insights into the cycling and preservation of DNA in environmental samples in the future. A general protocol is outlined, along with recommendations for optimizing this general protocol for specific sample types and research goals. PMID:26042110

  15. One-Dimensional Analysis of Thermal Stratification in AHTR and SFR Coolant Pools

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Per F. Peterson

    2007-10-01

    Thermal stratification phenomena are very common in pool type reactor systems, such as the liquid-salt cooled Advanced High Temperature Reactor (AHTR) and liquid-metal cooled fast reactor systems such as the Sodium Fast Reactor (SFR). It is important to accurately predict the temperature and density distributions both for design optimation and accident analysis. Current major reactor system analysis codes such as RELAP5 (for LWR’s, and recently extended to analyze high temperature reactors), TRAC (for LWR’s), and SASSYS (for liquid metal fast reactors) only provide lumped-volume based models which can only give very approximate results and can only handle simple cases with one mixing source. While 2-D or 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, yet such fine grid resolution is difficult or impossible to provide for studying the reactor response to transients due to computational expense. Therefore, new methods are needed to support design optimization and safety analysis of Generation IV pool type reactor systems. Previous scaling has shown that stratified mixing processes in large stably stratified enclosures can be described using one-dimensional differential equations, with the vertical transport by free and wall jets modeled using standard integral techniques. This allows very large reductions in computational effort compared to three-dimensional numerical modeling of turbulent mixing in large enclosures. The BMIX++ (Berkeley mechanistic MIXing code in C++) code was originally developed at UC Berkeley to implement such ideas. This code solves mixing and heat transfer problems in stably stratified enclosures. The code uses a Lagrangian approach to solve 1-D transient governing equations for the ambient fluid and uses analytical or 1-D integral models to compute substructures. By including liquid salt properties, BMIX++ code is

  16. Control of fermentation types in continuous-flow acidogenic reactors: effects of pH and redox potential

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The experiments were carried out in continuous-flow acidogenic reactors with molasses used as sub strate to study the effects of pH and redox potential on fermentation types. The conditions for each fermentation type were investigated at different experimental stages of start-up, pH-regulating and redox potential-regulating.The experiments confirmed that butyric acid-type fermentation would occur at pH > 6, the propionic acid-type fermentation at pH about 5.5 with Eh > - 278 mV, and the ethanol-type fermentation at pH < 4.5. A higher redox potential will lead to propionic acid-type fermentation because propionogens are facultative anaerobic bacteria.

  17. Criticality safety of the ET-RR-1 new spent fuel storage pool

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, E.; Sallam, O.H.; Amin, E

    2001-03-01

    A new ET-RR-1 spent fuel storage pool is now under construction on the reactor site at Inshass. In addition, the pool is designed to accommodate spent fuel of MTR type as well. Criticality safety of this pool for the different fuel types has been evaluated as a function of U{sup 235} loading. The effect of fuel element separation (rows and columns) on the eigenvalue has been studied. As a conservative assumption, the pool is assumed to be filled with fresh fuel. The eigenvalue considering a realistic degree of fuel burn-up was determined in order to determine the safety margin. The calculations have been carried out using the code packages of the National Center for Nuclear Safety and Radiation Control.

  18. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  19. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  20. Multi purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: vkrain@magnum.barc.ernet.in; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-04-15

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  1. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  2. A note on the evaluation of the guest-gas uptake into a clathrate hydrate being formed in a semibatch- or batch-type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Yasuhiko H.; Komae, Naoya [Department of Mechanical Engineering, Keio University, 3-14-1 Hiyoshi, Kohoku-ku, Yokohama 223-8522 (Japan)

    2008-05-15

    This paper deals with the principle of determining the rate of guest-gas uptake into a clathrate hydrate being formed in a semibatch-type isobaric reactor or a batch-type closed reactor on the basis of experimental data for the guest-gas supply into the reactor or the pressure change inside the reactor. The specific issue considered here is the possible necessity of taking into account the effect of the change in the total volume of the condensed (liquid + hydrate) phases inside the reactor during each hydrate-forming operation. General schemes for evaluating this effect in semibatch and batch operations are formulated and applied to some specific hydrate-forming operations to evaluate the effect on estimating the guest-gas uptake into the hydrate. (author)

  3. Experimental Study on the Thermal Stratification in a Pool Boiling with a Horizontal Heat Source

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Ryu, Sung Uk; Euh, Dong-Jin; Song, Chul-Hwa [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Thermal stratification is formed in horizontal fluid layers with different temperatures, where the warmer fluid layers are situated above the cooler fluid layers. Thermal stratification phenomena are common in pool type reactor systems, such as the liquid-salt cooled advanced high temperature reactor (AHTR) and liquid-metal cooled fast reactor systems such as the sodium fast reactor (SFR). Thermal stratification is increasingly encountered in large pools that are being used as heat sinks in the new generation of advanced reactors. The small-scale pool test was conducted to investigate the thermal stratification phenomena that occurred during the heat-up of a water in a pool. Because turbulence and boiling models affect the natural convection significantly, it is important to obtain local information regarding the fluid velocity and void distribution to determine the relevant physical models. To understand the flow phenomena inside a pool, a non-intrusive technique is adopted to measure the flow velocity field. In this study, the 2D particle image velocimetry (PIV) measurement technique is used to determine the fluid velocity vector field of single- and/or two-phase natural convection flow and thermal stratification in a pool. Detailed velocity measurements using the 2D PIV measurement technique were conducted to investigate single- and/or two-phase natural convection flow and thermal stratification in a pool boiling. In this study, the two-dimensional velocity vector fields as the water temperature increased were experimentally acquired in a pool that contained a horizontal heater rod. The experimental results indicate a large natural convection flow at the region above the heater rod and thermal stratification at the region below the heater rod. The flow of the opposite direction to each other was shown in the region between the heater rod and the thermal boundary layer. This flow pattern will contribute to maintain the thermal stratification and retard the water

  4. Application of a novel type impinging streams reactor in solid-liquid enzyme reactions and modeling of residence time distribution using GDB model.

    Science.gov (United States)

    Fatourehchi, Niloufar; Sohrabi, Morteza; Dabir, Bahram; Royaee, Sayed Javid; Haji Malayeri, Adel

    2014-02-01

    Solid-liquid enzyme reactions constitute important processes in biochemical industries. The isomerization of d-glucose to d-fructose, using the immobilized glucose isomerase (Sweetzyme T), as a typical example of solid-liquid catalyzed reactions has been carried out in one stage and multi-stage novel type of impinging streams reactors. Response surface methodology was applied to determine the effects of certain pertinent parameters of the process namely axial velocity (A), feed concentration (B), nozzles' flow rates (C) and enzyme loading (D) on the performance of the apparatus. The results obtained from the conversion of glucose in this reactor were much higher than those expected in conventional reactors, while residence time was decreased dramatically. Residence time distribution (RTD) in a one-stage impinging streams reactor was investigated using colored solution as the tracer. The results showed that the flow pattern in the reactor was close to that in a continuous stirred tank reactor (CSTR). Based on the analysis of flow region in the reactor, gamma distribution model with bypass (GDB) was applied to study the RTD of the reactor. The results indicated that RTD in the impinging streams reactor could be described by the latter model.

  5. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2017-05-15

    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  6. Safety and tolerability of sitagliptin in clinical studies: a pooled analysis of data from 10,246 patients with type 2 diabetes

    Directory of Open Access Journals (Sweden)

    Guo Hua

    2010-04-01

    Full Text Available Abstract Background In a previous pooled analysis of 12 double-blind clinical studies that included data on 6,139 patients with type 2 diabetes, treatment with sitagliptin, a dipeptidyl peptidase-4 (DPP-4 inhibitor, was shown to be generally well tolerated compared with treatment with control agents. As clinical development of sitagliptin continues, additional studies have been completed, and more patients have been exposed to sitagliptin. The purpose of the present analysis is to update the safety and tolerability assessment of sitagliptin by pooling data from 19 double-blind clinical studies. Methods The present analysis included data from 10,246 patients with type 2 diabetes who received either sitagliptin 100 mg/day (N = 5,429; sitagliptin group or a comparator agent (placebo or an active comparator (N = 4,817; non-exposed group. The 19 studies from which this pooled population was drawn represent the double-blind, randomized studies that included patients treated with the usual clinical dose of sitagliptin (100 mg/day for between 12 weeks and 2 years and for which results were available as of July 2009. These 19 studies assessed sitagliptin taken as monotherapy, initial combination therapy with metformin or pioglitazone, or as add-on combination therapy with other antihyperglycemic agents (metformin, pioglitazone, a sulfonylurea ± metformin, insulin ± metformin, or rosiglitazone + metformin. Patients in the non-exposed group were taking placebo, metformin, pioglitazone, a sulfonylurea ± metformin, insulin ± metformin, or rosiglitazone + metformin. The analysis used patient-level data from each study to evaluate between-group differences in the exposure-adjusted incidence rates of adverse events. Results Summary measures of overall adverse events were similar in the sitagliptin and non-exposed groups, except for an increased incidence of drug-related adverse events in the non-exposed group. Incidence rates of specific adverse events were

  7. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  8. DeveIopment of interactive safety anaIysis program for pooI type sodium cooIed fast reactor%池式钠冷快堆交互式安全分析软件开发

    Institute of Scientific and Technical Information of China (English)

    钱鸿涛; 李政昕; 胡文军; 宫宇

    2015-01-01

    为建立适用于池式钠冷快堆的仿真机,开发了基于法国快堆系统分析程序 OASIS 的交互式安全分析系统,实现了实时绘图、动态显示等可视化功能。利用该系统模拟了中国实验快堆的堆芯、主热传输系统、事故余热排出系统,以及控制调节系统和保护系统,分析了各个功率台阶的稳态及满功率下流量阶跃瞬态工况。分析结果与设计值符合度良好,表明该系统具有良好的适用性,可用于人员培训与安全审评等。%An interactive safety analysis program was developed and integrated into the simulation system for pool type sodium cooled fast reactor based on a French fast reactor system analysis code OASIS.The visualized functions of real-time plotting and dynamic display were provided.The core,main power transfer system,decay heat removal system,control and regulation system and reactor protection system of China Experimental Fast Reactor were simulated by the system.The various power level steady states and the transient of flow step at full power state were analyzed.The calculation results match well with the design data.It can be indicated that the program had a good applicability,and can be used for personnel training and safety review.

  9. Theoretical and Experimental Evaluation of the Temperature Distribution in a Dry Type Air Core Smoothing Reactor of HVDC Station

    Directory of Open Access Journals (Sweden)

    Yu Wang

    2017-05-01

    Full Text Available The outdoor ultra-high voltage (UHV dry-type air-core smoothing reactors (DASR of High Voltage Direct Current systems are equipped with a rain cover and an acoustic enclosure. To study the convective heat transfer between the DASR and the surrounding air, this paper presents a coupled model of the temperature and fluid field based on the structural features and cooling manner. The resistive losses of encapsulations calculated by finite element method (FEM were used as heat sources in the thermal analysis. The steady fluid and thermal field of the 3-D reactor model were solved by the finite volume method (FVM, and the temperature distribution characteristics of the reactor were obtained. Subsequently, the axial and radial temperature distributions of encapsulation were investigated separately. Finally, an optical fiber temperature measurement scheme was used for an UHV DASR under natural convection conditions. Comparative analysis showed that the simulation results are in good agreement with the experimental data, which verifies the rationality and accuracy of the numerical calculation. These results can serve as a reference for the optimal design and maintenance of UHV DASRs.

  10. Designing an epithermal neutron beam for boron neutron capture therapy for a DIDO type reactor using MCNP

    Science.gov (United States)

    Ross, D.; Constantine, G.; Weaver, D. R.; Beynon, T. D.

    1993-10-01

    This paper describes work undertaken to design an epithermal neutron beam for a DIDO type reactor for use in boron neutron capture therapy, a form of cancer treatment. It involved extensive use of MCNP, a Monte Carlo computer code. Initially, calculations were made with MCNP to simulate earlier experiments with an epithermal beam on the DIDO reactor. This comparison made it possible both to validate the Monte Carlo modelling of the reactor and to gain an insight into the important features of the simulation. Following this, MCNP was used to design a filtered epithermal neutron beam facility for DIDO's largest beam tube, a 13.7 cm radius horizontal tube which extends radially away from the core. First a selection was made of the optimum filter components for the beam. Then the research concentrated on combining these filter elements to construct a practical epithermal beam design. The results suggest that the optimum method of generating the epithermal neutron source is to employ a filter combination consisting principally of liquid argon with the addition of cadmium, aluminium, titanium and possibly tin. The calculations also show that the resultant neutron beam would have a flux greater than 1.0 × 10 9 n cm -2 s -1 and have sufficiently low fast-neutron and gamma-ray contamination.

  11. Design and analysis of 19 pin annular fuel rod cluster for pressure tube type boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Deokule, A.P., E-mail: abhijit.deokule1986@gmail.com [Homi Bhabha National Institute, Trombay 400 085, Mumbai (India); Vishnoi, A.K.; Dasgupta, A.; Umasankari, K.; Chandraker, D.K.; Vijayan, P.K. [Bhabha Atomic Research Centre, Trombay 400 085, Mumbai (India)

    2014-09-15

    Highlights: • Development of 19 pin annular fuel rod cluster. • Reactor physics study of designed annular fuel rod cluster. • Thermal hydraulic study of annular fuel rod cluster. - Abstract: An assessment of 33 pin annular fuel rod cluster has been carried out previously for possible use in a pressure tube type boiling water reactor. Despite the benefits such as negative coolant void reactivity and larger heat transfer area, the 33 pin annular fuel rod cluster is having lower discharge burn up as compared to solid fuel rod cluster when all other parameters are kept the same. The power rating of this design cannot be increased beyond 20% of the corresponding solid fuel rod cluster. The limitation on the power is not due to physics parameters rather it comes from the thermal hydraulics side. In order to increase power rating of the annular fuel cluster, keeping same pressure tube diameter, the pin diameter was increased, achieving larger inside flow area. However, this reduces the number of annular fuel rods. In spite of this, the power of the annular fuel cluster can be increased by 30% compared to the solid fuel rod cluster. This makes the nineteen pin annular fuel rod cluster a suitable option to extract more power without any major changes in the existing design of the fuel. In the present study reactor physics and thermal hydraulic analysis carried out with different annular fuel rod cluster geometry is reported in detail.

  12. Development of a resonant-type microwave reactor and its application to the synthesis of positron emission tomography radiopharmaceuticals.

    Science.gov (United States)

    Kimura, Hiroyuki; Yagi, Yusuke; Ohneda, Noriyuki; Odajima, Hiro; Ono, Masahiro; Saji, Hideo

    2014-10-01

    Microwave technology has been successfully applied to enhance the effectiveness of radiolabeling reactions. The use of a microwave as a source of heat energy can allow chemical reactions to proceed over much shorter reaction times and in higher yields than they would do under conventional thermal conditions. A microwave reactor developed by Resonance Instrument Inc. (Model 520/521) and CEM (PETWave) has been used exclusively for the synthesis of radiolabeled agents for positron emission tomography by numerous groups throughout the world. In this study, we have developed a novel resonant-type microwave reactor powered by a solid-state device and confirmed that this system can focus microwave power on a small amount of reaction solution. Furthermore, we have demonstrated the rapid and facile radiosynthesis of 16α-[(18)F]fluoroestradiol, 4-[(18)F]fluoro-N-[2-(1-methoxyphenyl)-1-piperazinyl]ethyl-N-2-pyridinylbenzamide, and N-succinimidyl 4-[(18)F]fluorobenzoate using our newly developed microwave reactor.

  13. Analysis by the Monte Carlo method of doses around the pool of storage of the control rods irradiated in a BWR reactor; Analisis mediante el metodo de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactror BWR

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J.; Gallardo, S.

    2011-07-01

    The control rods of a boiling water reactor (BWR) are subject to a neutron flux and thus become activated during their stay in the reactor core. Activation occurs especially in the stainless steel components and impurities. The activity generated results in a dose around the bar, while it le unimportant in the reactor, but to be taken into account when removed f ron it. The bars drawn are stored on hangers placed in the storage pools of spent fuel f ron the plant. Each hanger 12 accommodates control rods and are arranged so that at least three meters of water abode the heads of the control rods. The dose received by potentially exposed workers who are in the vicinity of the storage must be calculated to ensure adequate protection of the came. This dose can be decreased significantly by changing the arrangement of the bars on hangers.

  14. Heat Transfer Calculation on Plate-Type Fuel Assembly of High Flux Research Reactor

    Directory of Open Access Journals (Sweden)

    Daxin Gong

    2015-01-01

    Full Text Available Heat transfer characteristics of fuel assemblies for a high flux research reactor with a neutron trap are numerically investigated in this study. Single-phase turbulence flow is calculated by a commercial code, FLUENT, where the computational objective covers standard and control fuel assemblies. The simulation is carried out with an inlet coolant velocity varying from 4.5 m/s to 7.5 m/s in hot assemblies. The results indicate that the cladding temperature is always lower than the saturation temperature in the calculated ranges. The temperature rise in the control fuel assembly is smaller than that of the standard fuel assembly. Additionally, the assembly with a hot spot is specially studied, and the safety of the research reactor is also approved.

  15. CARBONACEOUS, NITROGENOUS AND PHOSPHORUS MATTERS REMOVAL FROM DOMESTIC WASTEWATER BY AN ACTIVATED SLUDGE REACTOR OF NITRIFICATION-DENITRIFICATION TYPE

    Directory of Open Access Journals (Sweden)

    MOHAMAD ALI FULAZZAKY

    2009-03-01

    Full Text Available This paper proposes an environmental engineering method based on biotechnology approach as one of the expected solutions that should be considered to implementing the activated sludge for improving the quality of water and living environment, especially to remove the major pollutant elements of domestic wastewater. Elimination of 3 major pollutant elements, i.e., carbon, nitrogen and phosphor containing the domestic wastewater is proposed to carry out biological method of an anoxic-aerobic reactor therein these types of pollutants should be consecutively processed in three steps. Firstly, eliminate the carbonaceous matter in the aerobic reactor. Secondly, to remove the carbonaceous and nitrogenous matters, it is necessary to modify the reactor’s nature from the aerobic condition to an anoxic-aerobic reactor. And finally, when the cycle of nitrification-denitrification is stable to achieve the target’s efficiency of reactor by adding the ferric iron into the activated sludge, it can be continued to remove the carbonaceous, nitrogenous and phosphorous matters simultaneously. The efficiency of carbonaceous and nitrogenous matters removal was confirmed with the effluent standard, COD is less than 100 mgO2/L and the value of global nitrogen is less than 10 mgN/L. The effectiveness of suspended matter removal is higher than 90% and the decantation of activated sludge is very good as identifying the Molhman’s index is below of 120 mL/L. The total phosphorus matter removal is more effective than the soluble phosphorus matter. By maintaining the reactor’s nature at the suitable condition, identifying the range of pH between 6.92 and 7.16 therefore the excellent abatement of phosphor of about 80% is achieving with the molar Fe/P ratio of 1.4.

  16. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

  17. Essential vernal pool habitat action plan

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — Vernal pool ecosystem conservation and recovery requires the recovery team to develop methods to determine the distribution of vernal pool types throughout the Great...

  18. Membrane reactor. Membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shindo, Y.; Wakabayashi, K. (National Chemical Laboratory for Industry, Tsukuba (Japan))

    1990-08-05

    Many reaction examples were introduced of membrane reactor, to be on the point of forming a new region in the field of chemical technology. It is a reactor to exhibit excellent function, by its being installed with membrane therein, and is generally classified into catalyst function type and reaction promotion type. What firstly belongs to the former is stabilized zirconia, where oxygen, supplied to the cathodic side of membrane with voltage, impressed thereon, becomes O {sup 2 {minus}} to be diffused through the membrane and supplied, as variously activated oxygenous species, on the anodic side. Examples with many advantages can be given such as methane coupling, propylene oxidation, methanating reaction of carbon dioxide, etc. Apart, palladium film and naphion film also belong to the former. While examples of the latter comprise, among others, decomposition of hydrogen sulfide by porous glass film and dehydrogenation of cyclohexane or palladium alloy film, which are expected to be developed and materialized in the industry. 33 refs., 8 figs.

  19. Electrochemical incineration of vinasse in filter-press-type FM01-LC reactor using 3D BDD electrode.

    Science.gov (United States)

    Nava, J L; Recéndiz, A; Acosta, J C; González, I

    2008-01-01

    This work shows results obtained in the electrochemical incineration of a synthetic vinasse with initial chemical oxygen demand (COD) of 75.096 g L(-1) in aqueous media (which resembles vinasse industrial wastewater). Electrolyses in a filter-press-type FM01-LC electrochemical reactor equipped with a three-dimensional (3D) boron doped diamond electrode (BDD) were performed at Reynolds values between 22 BDD surface. Experimental data revealed that hydrodynamic conditions slightly influence the vinasse degradation rate and current efficiency, indicating that the oxidation involves a complex pathway. IWA Publishing 2008.

  20. RCCS Experiments and Validation for High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Cliff Davis; Goon C. Park

    2007-09-01

    A reactor cavity cooling system (RCCS), an air-cooled helical coil RCCS unit immersed in the water pool, was proposed to overcome the disadvantages of the weak cooling ability of air-cooled RCCS and the complex structure of water-cooled RCCS for the high temperature gas-cooled reactor (HTGR). An experimental apparatus was constructed to investigate the various heat transfer phenomena in the water pool type RCCS, such as the natural convection of air inside the cavity, radiation in the cavity, the natural convection of water in the water pool and the forced convection of air in the cooling pipe. The RCCS experimental results were compared with published correlations. The CFX code was validated using data from the air-cooled portion of the RCCS. The RELAP5 code was validated using measured temperatures from the reactor vessel and cavity walls.

  1. Fatigue crack growth characteristics of nitrogen-alloyed type 347 stainless under operating conditions of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki Deuk; Hong, Seok Min; Kim, Dae Whan; Lee, Bong Sang [Korea Atomic Energy Research Institute, Nuclear Materials Safety Research Division, Daejeon (Korea, Republic of); Kim, Seon Jin [Hanyang University, Division of materials science and engineering, Seoul (Korea, Republic of)

    2017-06-15

    The fatigue crack growth behavior of Type 347 (S347) and Type 347N (S347N) stainless steel was evaluated under the operating conditions of a pressurized water reactor (PWR). These two materials showed different fatigue crack growth rates (FCGRs) according to the changes in dissolved oxygen content and frequency. Under the simulated PWR conditions for normal operation, the FCGR of S347N was lower than that of S347 and insensitive to the changes in PWR water conditions. The higher yield strength and better corrosion resistance of the nitrogen-alloyed Type 347 stainless steel might be a main cause of slower FCGR and more stable properties against changes in environmental conditions.

  2. Assessment of carbon pools in two soils from the Campania region (Southwest, Italy) under different forest types

    Science.gov (United States)

    Álvarez-Romero, Marta; Papa, Stefania; Lozano-García, Beatriz; Parras-Alcántara, Luis; González-Pérez, José A.; Jordán, Antonio; Zavala, Lorena M.; González-Vila, Francisco J.; Coppola, Elio

    2014-05-01

    suitable for the formation and accumulation of SOM in depth (Bh horizon) were studied. The content of the different soil C fractions was assessed for each soil profile and included: total extractable C, (TEC), total organic C (TOC), total extractable lipds (TEL), humified C (humic and fulvic acids, HA & FA) and non humic C (NHC), lignin C, cellulose C. Also were calculated parameters of humification, humification degree (DH), humification rate (HR), total level of humification (HU) and humification index (HI) The results are discussed in terms of how soil use and vegetation influences the identified C pools, and the humification indexes.

  3. Effect of application rates and media types on nitrogen and surfactant removal in trickling filters applied to the post-treatment of effluents from UASB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, P. G. S. de; Taveres, F. v. F.; Chernicharo, C. A. I.

    2009-07-01

    Tricking filters are a very promising alternative for the post treatment of effluents from UASB reactors treating domestic sewage,especially in developing countries. Although a fair amount of information is already available regarding organic mater removal in this combined system, very little is known in relation to nitrogen and surfactant removal in trickling filters post-UASB reactors. Therefore, the purpose of this study was to evaluate and compare the effect evaluate and compare the effect of different application rates and packing media types on trickling filters applied to the post-treatment of effluents from UASB reactors, regarding the removal of ammonia nitrogen and surfactants. (Author)

  4. Determination of the exposition rapidity in the level 49.90 of the reactor building for the decrease in the water level of the spent fuel pool; Determinacion de la rapidez de exposion en el nivel 49.90 del edificio del reactor por la disminucion en el nivel de agua de la alberca de combustible gastado

    Energy Technology Data Exchange (ETDEWEB)

    Mijangos D, Z. E.; Herrera H, S. F.; Cruz G, M. A.; Amador C, C., E-mail: zoedelfin@gmail.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Subgerencia de Ingenieria, Km 44.5 Carretera Cardel-Nautla, 91476 Laguna Verde, Alto Lucero, Veracruz (Mexico)

    2014-10-15

    The fuel assemblies storage in the nuclear power plant of Laguna Verde (NPP-L V) represents a crucial aspect, due to the generated dose by the decay heat of the present radio-nuclides in the assemblies retired of the reactor core, after their useful life. These spent assemblies are located inside the spent fuel pool (SFP), in the level 49.90 m in the Reload Floor of the Reactor building of NPP-L V. This leads to the protection at personnel applying the ALARA (As Low As Reasonably Achievable) criteria, fulfilling the established dose criteria by the Regulator Body the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). Considering the loss scenario of the cooling system of the SFP, in which the SFP water vaporizes, is important to know the water level in which the limit of effective dose equivalent is fulfilled for the personnel. Also, is important for the instrumentation of the SFP, for the useful life of the same instruments. In this work is obtained the exposition rapidity corresponding to different water levels of SFP in the Reload Floor of NPP-L V, to identify the minimum level of water where the limit of effective dose equivalent is fulfilled of 25 rem s to the personnel, established in the Article 48 of the General Regulation of Radiological Safety of CNSNS and the Chapter 50 Section 67 of the 10-Cfr of Nuclear Regulatory Commission in USA. The water level is also identified where the exposition rapidity is of 15 m R/hr, being the value of the set point of the area radiation monitor D21-Re-N003-1, located to 125 cm over the level 49.90 meters of the Reload Floor of NPP-L V. (Author)

  5. Analysis on the `Thermite` reaction consequences in accidents involving research reactors using plate-type fuel; Analisis sobre las concequencias de la reaccion `Termita` en caso de accidentes en reactores de investigacion que utilizan combustible tipo placa

    Energy Technology Data Exchange (ETDEWEB)

    Boero, Norma L.; Bruno, Hernan R.; Camacho, Esteban F.; Cincotta, Daniel O.; Yorio, Daniel [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Constituyentes

    1999-11-01

    The mixture of Al-U{sub 3} O{sub 8} is not in a state of chemical equilibrium, and at temperatures of between 850 deg C and 1000 deg C, it reacts exo thermally. This is known, in corresponding bibliography as a `Thermite reaction. This mixture is used in the manufacturing of the plate-type fuel used in research reactors. It has been pointed out that the release of energy caused by this type of reactions might represent a risk in case of accidents in this type of reactor. Conclusions, in general, tend to indicate that no such risk exists, although no concrete assurance is given that this is the case, and this fact, therefore, leaves room for doubt. The objective of this paper is to provide an in-depth study of what happens to a fuel plate when it is subjected to thermite reaction. We will, furthermore, analyze the consequences of the release of energy generated by this type of reaction within the core of the reactor, clearly defining the problem for this type of fuel and this kind of reactor. (author) 3 refs., 9 figs., 1 tab.

  6. Analysis on Electromagnetic Characteristics of Research Reactor Control Rod Drive Mechanism for Thrust Force Improvement

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Choi, Myoung Hwan; Yu, Je Yong; Cho, Yeong Garp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The control rod drive mechanism (CRDM) is the part of reactor regulating system (RRS), which is located in the reactor pool top or the room below the reactor pool. The function of the CRDM is to insert, withdraw or maintain neutron absorbing material (control rod) at any required position within the reactor core, in order to the reactivity of the core. There are so many kinds of CRDM, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we are investigating the movable coil electromagnetic drive mechanism (MCEDM) which is new scheme for the reactor control rod adopted by China Advanced Research Reactor (CARR). To have a better knowledge of the electromagnetic and magnetic characteristics, numerical models of MCEDM are proposed. Especially in order to achieve improved thrust force, numerical magnetic field calculations for various kinds of magnetic and electromagnetic configuration have been performed. As a result, we present the improved design of MCEDM for research reactor

  7. Design and installation of a hot water layer system at the Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila

    2013-01-01

    Full Text Available A hot water layer system (HWLS is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

  8. Bioregeneration of perchlorate-laden gel-type anion-exchange resin in a fluidized bed reactor.

    Science.gov (United States)

    Venkatesan, Arjun K; Sharbatmaleki, Mohamadali; Batista, Jacimaria R

    2010-05-15

    Selective ion-exchange resins are very effective to remove perchlorate from contaminated waters. However, these resins are currently incinerated after one time use, making the ion-exchange process incomplete and unsustainable for perchlorate removal. Resin bioregeneration is a new concept that combines ion-exchange with biological reduction by directly contacting perchlorate-laden resins with a perchlorate-reducing bacterial culture. In this research, feasibility of the bioregeneration of perchlorate-laden gel-type anion-exchange resin was investigated. Bench-scale bioregeneration experiments, using a fluidized bed reactor and a bioreactor, were performed to evaluate the feasibility of the process and to gain insight into potential mechanisms that control the process. The results of the bioregeneration tests suggested that the initial phase of the bioregeneration process might be controlled by kinetics, while the later phase seems to be controlled by diffusion. Feasibility study showed that direct bioregeneration of gel-type resin was effective in a fluidized-bed reactor, and that the resin could be defouled, reused, and repeatedly regenerated using the method applied in this research.

  9. Bioregeneration of perchlorate-laden gel-type anion-exchange resin in a fluidized bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Venkatesan, Arjun K.; Sharbatmaleki, Mohamadali [Department of Civil and Environmental Engineering, University of Nevada Las Vegas (UNLV), 4505 Maryland Parkway, Las Vegas, NV 89154-4015 (United States); Batista, Jacimaria R., E-mail: jaci@ce.unlv.edu [Department of Civil and Environmental Engineering, University of Nevada Las Vegas (UNLV), 4505 Maryland Parkway, Las Vegas, NV 89154-4015 (United States)

    2010-05-15

    Selective ion-exchange resins are very effective to remove perchlorate from contaminated waters. However, these resins are currently incinerated after one time use, making the ion-exchange process incomplete and unsustainable for perchlorate removal. Resin bioregeneration is a new concept that combines ion-exchange with biological reduction by directly contacting perchlorate-laden resins with a perchlorate-reducing bacterial culture. In this research, feasibility of the bioregeneration of perchlorate-laden gel-type anion-exchange resin was investigated. Bench-scale bioregeneration experiments, using a fluidized bed reactor and a bioreactor, were performed to evaluate the feasibility of the process and to gain insight into potential mechanisms that control the process. The results of the bioregeneration tests suggested that the initial phase of the bioregeneration process might be controlled by kinetics, while the later phase seems to be controlled by diffusion. Feasibility study showed that direct bioregeneration of gel-type resin was effective in a fluidized-bed reactor, and that the resin could be defouled, reused, and repeatedly regenerated using the method applied in this research.

  10. Single Phase Natural Circulation Behaviors of the Integral Type Marine Reactor Simulator under Rolling Motion Condition

    Directory of Open Access Journals (Sweden)

    Hou-jun Gong

    2015-01-01

    Full Text Available During operation in the sea the reactor natural circulation behaviors are affected by ship rolling motion. The development of an analysis code and the natural circulation behaviors of a reactor simulator under rolling motion are described in this paper. In the case of rolling motion, the primary coolant flow rates in the hot legs and heating channels oscillated periodically, and the amplitude of flow rate oscillation was in direct proportion to rolling amplitude, but in inverse proportion to rolling period. The total mass flow rate also oscillated with half the rolling period, and the average total mass flow rate was less than that in steady state. In the natural circulation under a rolling motion, the flow rate oscillations in the hot legs were controlled by the tangential force; however, the mass flow rate oscillations in the total natural circulation and the heating channels were a result of the combined action of the change of inclination angle, flow resistance, and the extra force arising from the rolling motion. The extra tangential force brought about intense flow rate oscillations in the hot legs, which resulted in increasing total flow resistance; however the extra centrifugal force played a role in increasing thermal driving head.

  11. Reactor pulse repeatability studies at the annular core research reactor

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

  12. Dissolved inorganic nitrogen pools and surface flux under different brackish marsh vegetation types, common reed (Phragmites australis) and salt hay (Spartina patens)

    Science.gov (United States)

    Windham-Myers, L.

    2005-01-01

    The current expansion of Phragmites australis into the high marsh shortgrass (Spartina patens, Distichlis spicata) communities of eastern U.S. salt marshes provided an opportunity to identify the influence of vegetation types on pools and fluxes of dissolved inorganic nitrogen (DIN). Two brackish tidal marshes of the National Estuarine Research Reserve system were examined, Piermont Marsh of the Hudson River NERR in New York and Hog Island in the Jacques Coustaeu NERR of New Jersey. Pools of DIN in porewater and rates of DIN surface flux were compared in replicated pairs of recently-expanded P. australis and neighboring S. patens-dominated patches on the high marsh surface. Both marshes generally imported nitrate (NO3-) and exported ammonium (NH4+), such that overall DIN was exported. No differences in surface exchange of NO3- or NH4+ were observed between vegetation types. Depth-averaged porewater NH4+ concentrations over the entire growing season were 56% lower under P. australis than under S. patens (average 1.4 vs. 3.2 mg NH4+ L-1) with the most profound differences in November. Porewater profiles showed an accumulation of NH4+ at depth in S. patens and constant low concentrations in P. australis from the soil surface to 50 cm depth, with no significant differences in porewater salinity. Despite these profound differences in porewater, NH 4+ diffusion from soils of P. australis and S. patens were not measurably different, were similar to other published rates, and were well below estimated rates based on passive diffusion alone. Rapid adsorption and uptake by litter and microbes in surface soils of both communities may buffer NH4+ loss to flooding tides in both communities, thereby reducing the impact of P. australis invasion on NH4+ flux to flooding waters. ?? Springer 2005.

  13. Dismantling design for the loop rooms on the MR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Craig, D.; Fecitt, L. [NUKEM Limited, Dounreay (United Kingdom); Gorlinsky, Yu.E. [RRC Kurchatov Institute, Moscow (Russian Federation); Harman, N.F.; Jackson, R. [Serco Technical and Assurance Services, Warrington (United Kingdom); Kolyadin, V.I. [RRC Kurchatov Institute, Moscow (Russian Federation); Lobach, Yu.N., E-mail: lobach@kinr.kiev.u [Institute for Nuclear Research of NASU, pr.Nauki, 47, 03680 Kiev (Ukraine); Pavlenko, V.I. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2009-12-15

    The recently completed international co-operation project was aimed at planning for decommissioning the MR reactor identified as a pilot plant for the decommissioning of the other shutdown reactors on the site. The MR reactor was a pool-type, materials testing reactor with the total thermal power of 50 MW which incorporated pressure tubes containing fuel under test. The MR facility includes the reactor with its nine loop rig rooms containing pumps, heat exchangers and experimental equipment as well as systems and equipment located in other buildings in the complex. The objective of the MR reactor decommissioning project was to identify dismantling equipment and the decommissioning methodology for the reactor, loop rooms and redundant services to permit the refit and re-use of the building for a different nuclear related purpose. The dismantling design comprises two separate, but combined, tasks, namely, the dismantling of reactor installation itself and dismantling of experimental loops. The techniques proposed to undertake the dismantling operations within the loop rooms are described. Two options have been developed for removing contaminated equipment from the high radiation field loop rooms and packaging the waste into approved waste containers. The benefits and detriments of both methods have been identified, which allows implementing the safe, timely and cost-effective decommissioning.

  14. Analysis of Task Types and Error Types of the Human Actions Involved in the Human-related Unplanned Reactor Trip Events

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Whan; Park, Jin Kyun; Jung, Won Dea

    2008-02-15

    This report provides the task types and error types involved in the unplanned reactor trip events that have occurred during 1986 - 2006. The events that were caused by the secondary system of the nuclear power plants amount to 67 %, and the remaining 33 % was by the primary system. The contribution of the activities of the plant personnel was identified as the following order: corrective maintenance (25.7 %), planned maintenance (22.8 %), planned operation (19.8 %), periodic preventive maintenance (14.9 %), response to a transient (9.9 %), and design/manufacturing/installation (9.9%). According to the analysis of error modes, the error modes such as control failure (22.2 %), wrong object (18.5 %), omission (14.8 %), wrong action (11.1 %), and inadequate (8.3 %) take up about 75 % of all the unplanned trip events. The analysis of the cognitive functions involved showed that the planning function makes the highest contribution to the human actions leading to unplanned reactor trips, and it is followed by the observation function (23.4%), the execution function (17.8 %), and the interpretation function (10.3 %). The results of this report are to be used as important bases for development of the error reduction measures or development of the error mode prediction system for the test and maintenance tasks in nuclear power plants.

  15. Hawaii ESI: POOLS (Anchialine Pool Points)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for anchialine pools in Hawaii. Anchialine pools are small, relatively shallow coastal ponds that occur...

  16. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  17. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    Science.gov (United States)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  18. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  19. Characterization of the biomass of a hybrid anaerobic reactor (HAR with two types of support material during the treatment of the coffee wastewater

    Directory of Open Access Journals (Sweden)

    Vivian Galdino da Silva

    2013-06-01

    Full Text Available This study investigated the microbiology of a hybrid anaerobic reactor (HAR in the removal of pollutant loads. This reactor had the same physical structure of an UASB reactor, however with minifilters inside containing two types of support material: expanded clay and gravel. Two hydraulic retention times (HRT of 24h and 18h were evaluated at steady-state conditions, resulting in organic loading rates (OLR of 0.032 and 0.018 kgDBO5m-3d-1 and biological organic loading rates (BOLR of 0,0015 and 0.001 kgDBO5kgSVT- 1d¹, respectively. The decrease in concentration of organic matter in the influent resulted an endogenous state of the biomass in the reactor. The expanded clay was the best support material for biofilm attachment.

  20. Influences of Excess Oscillation of Voltage Pulse and Discharge Mode on NO Removal Using Barrier-Type Plasma Reactor

    Science.gov (United States)

    Kadowaki, Kazunori; Suzuki, Yoshiaki; Ihori, Haruo; Kitani, Isamu

    This paper presents experimental results of NO removal from a simulated exhausted-gas using a barrier type reactor with screw electrodes subjected to polarity-reversed voltage pulses. The polarity-reversed pulse was produced by direct grounding of a charged coaxial cable because a traveling wave voltage was negatively reflected at the grounding end with a change in its polarity and then it propagated to the plasma reactor at the opposite end. Influence of cable length on NO removal was studied for two kinds of cable connection, single-connected cable and parallel-connected cables. NO removal ratio for a 50m-long cable was lower than that for much shorter cables in both single and parallel connections when the applied voltage became high. Energy efficiency for NO removal also increased with decreasing the cable length. This was because excess discharges during the voltage oscillation caused by the large stored energy in the long cable resulted in reproduction of NO molecules. Energy efficiency was further improved by changing the discharge mode from dielectric barrier discharge (DBD) to surface discharge (SD). Energy efficiency was up to 110g/kWh with 55% NO removal ratio and 34g/kWh with 100% NO removal ratio by using a single 10m-long cable in SD mode.

  1. Environmental impact assessment of a package type IFAS reactor during construction and operational phases: a life cycle approach.

    Science.gov (United States)

    Singh, Nitin Kumar; Singh, Rana Pratap; Kazmi, Absar Ahmad

    2017-05-01

    In the present study, a life cycle assessment (LCA) approach was used to analyse the environmental impacts associated with the construction and operational phases of an integrated fixed-film activated sludge (IFAS) reactor treating municipal wastewater. This study was conducted within the boundaries of a research project that aimed to investigate the implementation related challenges of a package type IFAS reactor from an environmental perspective. Along with the LCA results of the construction phase, a comparison of the LCA results of seven operational phases is also presented in this study. The results showed that among all the inputs, the use of stainless steel in the construction phase caused the highest impact on environment, followed by electricity consumption in raw materials production. The impact of the construction phase on toxicity impact indicators was found to be significant compared to all operational phases. Among the seven operational phases of this study, the dissolved oxygen phase III, having a concentration of ∼4.5 mg/L, showed the highest impact on abiotic depletion, acidification, global warming, ozone layer depletion, human toxicity, fresh water eco-toxicity, marine aquatic eco-toxicity, terrestrial eco-toxicity, and photochemical oxidation. However, better effluent quality in this phase reduced the eutrophication load on environment.

  2. Fuel shuffling optimization for the Delft research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft Univ. of Technology, Interfaculty Reactor Inst., Delft (Netherlands); Quist, A.J. [Delft Univ., Fac. of Applied Mathematics and Informatics, Delft (Netherlands)

    1997-07-01

    A fuel shuffling optimization procedure is proposed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, a 2 MWth swimming-pool type research reactor. In order to cope with the fluctuatory behaviour of objective functions in loading pattern optimization, the proposed cyclic permutation optimization procedure features a gradual transition from global to local search behaviour via the introduction of stochastic tests for the number of fuel assemblies involved in a cyclic permutation. The possible objectives and the safety and operation constraints, as well as the optimization procedure, are discussed, followed by some optimization results for the HOR. (author)

  3. Applied methods for mitigation of damage by stress corrosion in BWR type reactors; Metodos aplicados para la mitigacion del dano por corrosion bajo esfuerzo en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez C, R.; Diaz S, A.; Gachuz M, M.; Arganis J, C. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The Boiling Water nuclear Reactors (BWR) have presented stress corrosion problems, mainly in components and pipes of the primary system, provoking negative impacts in the performance of energy generator plants, as well as the increasing in the radiation exposure to personnel involucred. This problem has caused development of research programs, which are guided to find solution alternatives for the phenomena control. Among results of greater relevance the control for the reactor water chemistry stands out particularly in the impurities concentration and oxidation of radiolysis products; as well as the supervision in the materials selection and the stresses levels reduction. The present work presents the methods which can be applied to diminish the problems of stress corrosion in BWR reactors. (Author)

  4. SCWO characteristics of organics in a vertical type continuous reactor; Renzokushiki tategata hannoki ni yoru yukibutsu no chorinaki suisanka kyodo

    Energy Technology Data Exchange (ETDEWEB)

    Sekikawa, R.M.; Usui, T.; Nishimura, T.; Sato, H.; Hamada, S.; Sekino, H. [Ebara Research Co., Kanagawa (Japan). Center for Advanced Research

    2000-01-10

    SCWO characteristics are investigated for a vertical type, down stream continuous reactor system with mixing nozzle and sapphire windows. 2-propanol, hexane and biphenyl solution are used as fuel and air as oxidizer. 2-propanol is observed to be effective as makeup fuel to keep a stable autogenic SCWO reaction. Even for low air ratio as 1.1, high decomposition rate without CO, NO, NO{sub 2} or soot production is achieved. Calculated and experimental flue gas composition is in good agreement for a wide range of air ratio. Spontaneous flame formation is observed for SCWO of 2-propanol using air ratio over 1.8. These flame formations are not particular to 2-propanol and are also confirmed when using hexane and biphenyl solution as fuel. (author)

  5. Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.

  6. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  7. The Design, Fabrication, and Characteristic Experiment of the Electromagnet of Bottom-mounted Control Rod Drive Mechanism for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Cho, Yeong Garp; Choi, Myoung Hwan; Kim, Ji Ho; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    A control rod drive mechanism (CRDM) is located in the reactor pool top (Top-mounted) or the room below the reactor pool (Bottom-mounted). The function of the CRDM is to insert, withdraw, or maintain neutron absorbing material at any required position in the reactor core in order to maintain reactivity control of the core. There are so many kinds of CRDMs, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we are investigating the bottom-mounted control rod drive mechanism as shown in Fig. 1. To have a better knowledge of the electromagnetic and magnetic characteristics, numerical models of bottom-mounted CEDM are investigated. In this study, we clarified thrust force characteristics of the electromagnet by experiment and simulation, and verified the propriety of the FEM analysis by comparing it with the results

  8. Evaluation of plate type fuel options for small power reactors; Avaliacao de alternativas de combustivel tipo placa para reatores de pequeno porte

    Energy Technology Data Exchange (ETDEWEB)

    Andrzejewski, Claudio de Sa

    2005-07-01

    Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO{sub 2} in stainless steel, of UO{sub 2} in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

  9. Assessment of susceptibility of Type 304 stainless steel to intergranular stress corrosion cracking in simulated Savannah River Reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Ondrejcin, R.S.; Caskey, C.R. Jr.

    1989-12-01

    Intergranular stress corrosion cracking (IGSCC) of Type 304 stainless steel rate tests (CERT) of specimens machined was evaluated by constant extension from Savannah River Plant (SRP) decontaminated process water piping. Results from 12 preliminary CERT tests verified that IGSCC occurred over a wide range of simulated SRP envirorments. 73 specimens were tested in two statistical experimental designs of the central composite class. In one design, testing was done in environments containing hydrogen peroxide; in the other design, hydrogen peroxide was omitted but oxygen was added to the environment. Prediction equations relating IGSCC to temperature and environmental variables were formulated. Temperature was the most important independent variable. IGSCC was severe at 100 to 120C and a threshold temperature between 40C and 55C was identified below which IGSCC did not occur. In environments containing hydrogen peroxide, as in SRP operation, a reduction in chloride concentration from 30 to 2 ppB also significantly reduced IGSCC. Reduction in sulfate concentration from 50 to 7 ppB was effective in reducing IGSCC provided the chloride concentration was 30 ppB or less and temperature was 95C or higher. Presence of hydrogen peroxide in the environment increased IGSCC except when chloride concentration was 11 ppB or less. Actual concentrations of hydrogen peroxide, oxygen and carbon dioxide did not affect IGSCC. Large positive ECP values (+450 to +750 mV Standard Hydrogen Electrode (SHE)) in simulated SRP environments containing hydrogen peroxide and were good agreement with ECP measurements made in SRP reactors, indicating that the simulated environments are representative of SRP reactor environments. Overall CERT results suggest that the most effective method to reduce IGSCC is to reduce chloride and sulfate concentrations.

  10. Assessment of susceptibility of Type 304 stainless steel to intergranular stress corrosion cracking in simulated Savannah River Reactor environments

    Energy Technology Data Exchange (ETDEWEB)

    Ondrejcin, R.S.; Caskey, C.R. Jr.

    1989-12-01

    Intergranular stress corrosion cracking (IGSCC) of Type 304 stainless steel rate tests (CERT) of specimens machined was evaluated by constant extension from Savannah River Plant (SRP) decontaminated process water piping. Results from 12 preliminary CERT tests verified that IGSCC occurred over a wide range of simulated SRP envirorments. 73 specimens were tested in two statistical experimental designs of the central composite class. In one design, testing was done in environments containing hydrogen peroxide; in the other design, hydrogen peroxide was omitted but oxygen was added to the environment. Prediction equations relating IGSCC to temperature and environmental variables were formulated. Temperature was the most important independent variable. IGSCC was severe at 100 to 120C and a threshold temperature between 40C and 55C was identified below which IGSCC did not occur. In environments containing hydrogen peroxide, as in SRP operation, a reduction in chloride concentration from 30 to 2 ppB also significantly reduced IGSCC. Reduction in sulfate concentration from 50 to 7 ppB was effective in reducing IGSCC provided the chloride concentration was 30 ppB or less and temperature was 95C or higher. Presence of hydrogen peroxide in the environment increased IGSCC except when chloride concentration was 11 ppB or less. Actual concentrations of hydrogen peroxide, oxygen and carbon dioxide did not affect IGSCC. Large positive ECP values (+450 to +750 mV Standard Hydrogen Electrode (SHE)) in simulated SRP environments containing hydrogen peroxide and were good agreement with ECP measurements made in SRP reactors, indicating that the simulated environments are representative of SRP reactor environments. Overall CERT results suggest that the most effective method to reduce IGSCC is to reduce chloride and sulfate concentrations.

  11. Actinide neutron induced cross section measurements using the oscillation technique in the Minerve reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, B.; Leconte, P.; Gruel, A.; Antony, M.; Di-Salvo, J.; Hudelot, J.P.; Pepino, A.; Lecluze, A. [CEA Cadarache, DEN/CAD/DER/SPRC/LEPh, 13 - Saint-Paul-lez-Durance (France)

    2009-07-01

    CEA is deeply involved research programs concerning nuclear fuel advanced studies (actinides, plutonium), waste management, the scientific and technical support of French PWR reactors and EPR reactor, and innovative systems. In this framework, specific neutron integral experiments have been carried out in the critical ZPR (zero power reactor) facilities of the CEA at Cadarache such as MINERVE, EOLE and MASURCA. This paper deals with MINERVE Pool Reactor experiments. MINERVE is mainly devoted to neutronics studies of different reactor core types. The aim is to improve the knowledge of the integral absorption cross sections of actinides (OSMOSE program), of new absorbers (OCEAN program) and also for fission Products (CBU program) in thermal, epithermal and fast neutron spectra. (authors)

  12. [Soil organic carbon pools and their turnover under two different types of forest in Xiao-xing'an Mountains, Northeast China].

    Science.gov (United States)

    Gao, Fei; Jiang, Hang; Cui, Xiao-yang

    2015-07-01

    Soil samples collected from virgin Korean pine forest and broad-leaved secondary forest in Xiaoxing'an Mountains, Northeast China were incubated in laboratory at different temperatures (8, 18 and 28 °C) for 160 days, and the data from the incubation experiment were fitted to a three-compartment, first-order kinetic model which separated soil organic carbon (SOC) into active, slow, and resistant carbon pools. Results showed that the soil organic carbon mineralization rates and the cumulative amount of C mineralized (all based on per unit of dry soil mass) of the broad-leaved secondary forest were both higher than that of the virgin Korean pine forest, whereas the mineralized C accounted for a relatively smaller part of SOC in the broad-leaved secondary forest soil. Soil active and slow carbon pools decreased with soil depth, while their proportions in SOC increased. Soil resistant carbon pool and its contribution to SOC were both greater in the broad-leaved secondary forest soil than in the virgin Korean pine forest soil, suggesting that the broad-leaved secondary forest soil organic carbon was relatively more stable. The mean retention time (MRT) of soil active carbon pool ranged from 9 to 24 d, decreasing with soil depth; while the MRT of slow carbon pool varied between 7 and 24 a, increasing with soil depth. Soil active carbon pool and its proportion in SOC increased linearly with incubation temperature, and consequently, decreased the slow carbon pool. Virgin Korean pine forest soils exhibited a higher increasing rate of active carbon pool along temperature gradient than the broad-leaved secondary forest soils, indicating that the organic carbon pool of virgin Korean pine forest soil was relatively more sensitive to temperature change.

  13. Research on Three-Phase Magnetic Valve Type Controllable Reactor%三相磁阀式可控电抗器的研究

    Institute of Scientific and Technical Information of China (English)

    李海洋; 赵国生

    2011-01-01

    提出了一种新型的三相磁阀式可控电抗器,并介绍了其结构及原理,对其进行了电磁分析。%The new three-phase magnetic valve type controllable reactor is presented.Its structure and principle are introduced.Its electromagnetic problems are analyzed.

  14. Poole-Frenkel effect on electrical characterization of Al-doped ZnO films deposited on p-type GaN

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Bohr-Ran [Graduate Institute of Electro-Optical Engineering and Department of Electronic Engineering, National Taiwan University of Science and Technology, Taipei 106, Taiwan (China); Liao, Chung-Chi [Department of Electronic Engineering, National Taiwan University of Science and Technology, Taipei 106, Taiwan (China); Ke, Wen-Cheng, E-mail: wcke@saturn.yzu.edu.tw; Chang, Yuan-Ching; Huang, Hao-Ping [Department of Mechanical Engineering, Yuan Ze University, Chung-Li 320, Taiwan (China); Chen, Nai-Chuan [Institute of Electro-Optical Engineering and Department of Electronic Engineering, Chang Gung University, Tao-Yuan 333, Taiwan (China)

    2014-03-21

    This paper presents the electrical properties of Al-doped ZnO (AZO) films directly grown on two types of p-type GaN thin films. The low-pressure p-GaN thin films (LP-p-GaN) exhibited structural properties of high-density edge-type threading dislocations (TDs) and compensated defects (i.e., nitrogen vacancy). Compared with high-pressure p-GaN thin films (HP-p-GaN), X-ray photoemission spectroscopy of Ga 3d core levels indicated that the surface Fermi-level shifted toward the higher binding-energy side by approximately 0.7 eV. The high-density edge-type TDs and compensated defects enabled surface Fermi-level shifting above the intrinsic Fermi-level, causing the surface of LP-p-GaN thin films to invert to n-type semiconductor. A highly nonlinear increase in leakage current regarding reverse-bias voltage was observed for AZO/LP-p-GaN. The theoretical fits for the reverse-bias voltage region indicated that the field-assisted thermal ionization of carriers from defect associated traps, which is known as the Poole-Frenkel effect, dominated the I-V behavior of AZO/LP-p-GaN. The fitting result estimated the trap energy level at 0.62 eV below the conduction band edge. In addition, the optical band gap increased from 3.50 eV for as-deposited AZO films to 3.62 eV for 300 °C annealed AZO films because of the increased carrier concentration. The increasing Fermi-level of the 300 °C annealed AZO films enabled the carrier transport to move across the interface into the LP-p-GaN thin films without any thermal activated energy. Thus, the Ohmic behavior of AZO contact can be achieved directly on the low-pressure p-GaN films at room temperature.

  15. Sonochemical Reactors.

    Science.gov (United States)

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  16. Implications of reactor type and conditions on first-order hydrolysis rate assessment of maize

    NARCIS (Netherlands)

    Pabon Pereira, C.P.; Zeeman, G.; Zhao, R.; Ekmekci, B.; Lier, van J.B.

    2009-01-01

    The biodegradability and first-order hydrolysis coefficient of maize silage have been assessed from batch experiments using different types of inoculum and substrate to inocula (S/I) ratios, and from CSTRs working at different hydraulic retention times (HRTs). In the batch experiments, the assessed

  17. Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; J. E. O' Brien; J. S. Herring

    2007-09-01

    This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor – power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

  18. Design of an equilibrium nucleus of a BWR type reactor based in a Thorium-Uranium fuel; Diseno de un nucleo de equilibrio de un reactor tipo BWR basado en un combustible de Torio-Uranio

    Energy Technology Data Exchange (ETDEWEB)

    Francois, J.L.; Nunez C, A. [Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Facultad de Ingenieria-UNAM, Paseo Cuauhnahuac 8532, Jiutepec, Morelos (Mexico)

    2003-07-01

    In this work the design of the reactor nucleus of boiling water using fuel of thorium-uranium is presented. Starting from an integral concept based in a type cover-seed assemble is carried out the design of an equilibrium reload for the nucleus of a reactor like that of the Laguna Verde Central and its are analyzed some of the main design variables like the cycle length, the reload fraction, the burnt fuel, the vacuum distribution, the generation of lineal heat, the margin of shutdown, as well as a first estimation of the fuel cost. The results show that it is feasible to obtain an equilibrium reload, comparable to those that are carried out in the Laguna Verde reactors, with a good behavior of those analyzed variables. The cost of the equilibrium reload designed with the thorium-uranium fuel is approximately 2% high that the uranium reload producing the same energy. It is concluded that it is convenient to include burnable poisons, type gadolinium, in the fuel with the end of improving the reload design, the fuel costs and the margin of shutdown. (Author)

  19. Bellows-Type Accumulators for Liquid Metal Loops of Space Reactor Power Systems

    Science.gov (United States)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2006-01-01

    In many space nuclear power systems, the primary and/or secondary loops use liquid metal working fluids, and require accumulators to accommodate the change in the liquid metal volume and maintain sufficient subcooling to avoid boiling. This paper developed redundant and light-weight bellows-type accumulators with and without a mechanical spring, and compared the operating condition and mass of the accumulators for different types of liquid metal working fluids and operating temperatures: potassium, NaK-78, sodium and lithium loops of a total capacity of 50 liters and nominal operating temperatures of 840 K, 860 K, 950 K and 1340 K, respectively. The effects of using a mechanical spring and different structural materials on the design, operation and mass of the accumulators are also investigated. The structure materials considered include SS-316, Hastelloy-X, C-103 and Mo-14Re. The accumulator without a mechanical spring weighs 23 kg and 40 kg for a coolant subcooling of 50 K and 100 K, respectively, following a loss of the fill gas. The addition of a mechanical spring comes with a mass penalty, in favor of higher redundancy and maintaining a higher liquid metal subcooling.

  20. Cigarette smoking and lung cancer--relative risk estimates for the major histological types from a pooled analysis of case-control studies.

    Science.gov (United States)

    Pesch, Beate; Kendzia, Benjamin; Gustavsson, Per; Jöckel, Karl-Heinz; Johnen, Georg; Pohlabeln, Hermann; Olsson, Ann; Ahrens, Wolfgang; Gross, Isabelle Mercedes; Brüske, Irene; Wichmann, Heinz-Erich; Merletti, Franco; Richiardi, Lorenzo; Simonato, Lorenzo; Fortes, Cristina; Siemiatycki, Jack; Parent, Marie-Elise; Consonni, Dario; Landi, Maria Teresa; Caporaso, Neil; Zaridze, David; Cassidy, Adrian; Szeszenia-Dabrowska, Neonila; Rudnai, Peter; Lissowska, Jolanta; Stücker, Isabelle; Fabianova, Eleonora; Dumitru, Rodica Stanescu; Bencko, Vladimir; Foretova, Lenka; Janout, Vladimir; Rudin, Charles M; Brennan, Paul; Boffetta, Paolo; Straif, Kurt; Brüning, Thomas

    2012-09-01

    Lung cancer is mainly caused by smoking, but the quantitative relations between smoking and histologic subtypes of lung cancer remain inconclusive. By using one of the largest lung cancer datasets ever assembled, we explored the impact of smoking on risks of the major cell types of lung cancer. This pooled analysis included 13,169 cases and 16,010 controls from Europe and Canada. Studies with population controls comprised 66.5% of the subjects. Adenocarcinoma (AdCa) was the most prevalent subtype in never smokers and in women. Squamous cell carcinoma (SqCC) predominated in male smokers. Age-adjusted odds ratios (ORs) were estimated with logistic regression. ORs were elevated for all metrics of exposure to cigarette smoke and were higher for SqCC and small cell lung cancer (SCLC) than for AdCa. Current male smokers with an average daily dose of >30 cigarettes had ORs of 103.5 (95% confidence interval (CI): 74.8-143.2) for SqCC, 111.3 (95% CI: 69.8-177.5) for SCLC and 21.9 (95% CI: 16.6-29.0) for AdCa. In women, the corresponding ORs were 62.7 (95% CI: 31.5-124.6), 108.6 (95% CI: 50.7-232.8) and 16.8 (95% CI: 9.2-30.6), respectively. Although ORs started to decline soon after quitting, they did not fully return to the baseline risk of never smokers even 35 years after cessation. The major result that smoking exerted a steeper risk gradient on SqCC and SCLC than on AdCa is in line with previous population data and biological understanding of lung cancer development.

  1. The Performance Test for Reactor Coolant Pump (RCP) adopting Variable Restriction Orifice Type Control Valve

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.; Bae, B. U.; Cho, Y. J. and others

    2014-05-15

    The design values of the RCPTF are 17.2 MPa, 343 .deg. C, 11.7 m{sup 3}/s, and 13 MW in the maximum pressure, temperature, flow rate, and electrical power, respectively. In the RCPTF, various types of tests can be performed including a hydraulic performance test to acquire a H-Q curve as well seal transient tests, thrust bearing transient test, cost down test, NPSHR verification test, and so on. After a commissioning startup test was successfully perfomed, mechanical structures are improved including a flow stabilizer and variable restriction orifice. Two- branch pipe (Y-branch) was installed to regulate the flow rate in the range of performance tests. In the main pipe, a flow restrictor (RO: Restriction Orifice) for limiting the maximum flow rate was installed. In the branch pipe line, a globe valve and a butterfly valves for regulating the flow rate was located on the each branch line. When the pressure loss of the valve side is smaller than that of the RO side, the flow rate of valve side was increasing and the flow disturbance was occurred in the lower pipe line. Due to flow disturbnace, it is to cause an error when measuring RCP head and flow measurement of the venturi flow meter installed in the lower main pipe line, and thus leading to a decrease in measurement accuracy as a result. To increase the efficiency of the flow control availability of the test facility, the variable restriction orifice (VRO) type flow control valve was designed and manufactured. In the RCPTF in KAERI, the performance tests and various kinds of transient tests of the RCP were successfully performed. In this study, H-Q curve of the pump using the VRO revealed a similar trend to the result from two ROs. The VRO was confirmed to effectively cover the full test range of the flow rate.

  2. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  3. Improvement of hydrogen production via ethanol-type fermentation in an anaerobic down-flow structured bed reactor.

    Science.gov (United States)

    Anzola-Rojas, Mélida del Pilar; Zaiat, Marcelo; De Wever, Heleen

    2016-02-01

    Although a novel anaerobic down-flow structured bed reactor has shown feasibility and stable performance for a long-term compared to other anaerobic fixed bed systems for continuous hydrogen production, the volumetric rates and yields have so far been too low. In order to improve the performance, an operation strategy was applied by organic loading rate (OLR) variation (12-96 g COD L(-1) d(-1)). Different volumetric hydrogen rates, and yields at the same OLR indicated that the system was mainly driven by the specific organic load (SOL). When SOL was kept between 3.8 and 6.2 g sucrose g(-1) VSS d(-1), the volumetric rates raised from 0.1 to 8.9 L H2 L(-1) d(-1), and the yields were stable around 2.0 mol H2 mol(-1) converted sucrose. Furthermore, hydrogen was produced mainly via ethanol-type fermentation, reaching a total energy conversion rate of 23.40 kJ h(-1) L(-1) based on both hydrogen and ethanol production.

  4. Scaled-up bioconversion of fish waste to liquid fertilizer using a 5 L ribbon-type reactor.

    Science.gov (United States)

    Dao, Van Thingoc; Kim, Joong Kyun

    2011-10-01

    A scaled-up conversion process of fish waste to liquid fertilizer was performed in a 5 L ribbon-type reactor. Biodegradation was performed by inoculation of autoclaved fish waste with 5.84 × 10(5) CFU mL(-1) of mixed microorganisms for 96 h. As a result, the pH changed from 6.92 to 5.72, the cell number reached 7.28 × 10(5) CFU mL(-1), and approximately 430 g (28.3%) of fish waste was degraded. Analyses indicated that the 96 h culture of inoculated fish waste possessed comparable fertilizing ability to commercial fertilizers in hydroponic culture with amino acid contents of 6.91 g 100 g(-1). Therefore, the scaled-up production achieved a more satisfactory fish waste degradation rate (3.61 g h(-1)) than the flask-scale production (0.24 g h(-1)). The biodegraded broth of fish waste at room temperature did not undergo putrefaction for 6 months due to the addition of 1% lactate.

  5. Microbial community changes during the start-up of an anaerobic/aerobic/anoxic-type sequencing batch reactor.

    Science.gov (United States)

    Zhang, Qian; He, Jiajie; Wang, Hongyu; Ma, Fang; Yang, Kai; Wang, Jingbo

    2013-01-01

    An anaerobic/aerobic/anoxic-type sequencing batch reactor was started up during a summer rainy season to obtain enhanced biological phosphorus removal (EBPR), and its sludge microbial community was also monitored in the hope of observing the microbial community evolution of polyphosphate-accumulating organisms (PAOs). During the start-up process, a total of 17 bands of highest species richness were detected in the sludge microbial community, including Alpha-, Beta-, and Gamma- Proteobacteria, as well as Actinobacteria and Planctomycetes. Major microbial community structural change was observed in Rhodocyclus-related and Acinetobacter-related PAOs, glycogen-accumulating organisms (GAOs), and Actinobacteria. In contrast to the current belief that enrichment of PAOs is essential for the establishment of EBPR, PAOs were not favourably enriched in this study. Instead, Actinobacteria and GAOs overwhelmingly flourished. The overall conclusion of this study challenges the conventional view that EBPR cannot live without traditional PAOs. However, it suggests an non-negligible role of denitrifying phosphorus-accumulating bacteria in EBPR systems, as well as other uncultured bacteria.

  6. Estimation of thermal loads on the VVER vessel under conditions of inversion of the stratified molten pool in a severe accident

    Science.gov (United States)

    Loktionov, V. D.; Mukhtarov, E. S.

    2016-09-01

    Analysis of the thermal state of molten pools that can be formed on the vessel bottom of the VVER-600 medium-power reactor during a severe anticipated accident with melting of the core is represented. Two types of the molten pool of core materials, with the two-layer and inverse three-layer stratification, are considered. Thermal loads acting on the reactor vessel from the melt are estimated depending on its formation time. Features of the thermal state of the melt in the case of its inverse stratification are analyzed. It is shown that thermal loads on the reactor vessel exceed the critical heat flux (CHF) when forming the two-layer stratified molten pool 10 and 24 h after its shutdown, and the thermal load is close to the corresponding CHF or somewhat exceeds it in 72 h. In the case of the formation of the inverse structure of the melt, one can observe a decrease by more than 2.5 times (in comparison with the two-layer stratified structure) in the thermal load on the reactor vessel in the region of its contact with the upper layer of the steel melt. Analysis of results showed that maximum densities of heat flux to the reactor vessel from the bottom metallic layer with the melt inversion did not exceed corresponding CHFs 24 and 72 h after the reactor shutdown. Because the thermal load on the reactor vessel can be localized in the region of its bottom, where the CHF is relatively small, during the inverse stratification of the melt, there is a need to carry out further in-depth experimental and analytical investigations of conditions for formation of the stratified molten pool and to obtain corrected experimental CHFs for conditions and outlines of cooling the external surface of the VVER-600 vessel in a severe accident.

  7. A study on the recriticality possibilities of fast reactor cores after a hypothetical core meltdown accident

    Energy Technology Data Exchange (ETDEWEB)

    Na, Byung Chan; Han, Do Hee; Kim, Young Cheol

    1997-04-01

    The preliminary and parametric sensitivity study on recriticality risk of fast reactor cores after a hypothetical total core meltdown accident was performed. Only the neutronic aspects of the accident was considered for this study, independent of the accident scenario. Estimation was made for the quantities of molten fuel which must be ejected out of the core in order to assure a sub-critical state. Diverse parameters were examined: molten pool type (homogenized or stratified), fuel temperature, conditions of the reactor core, core size (small or large), and fuel type (oxide, nitride, metal) (author). 7 refs.

  8. Are fast explorers slow reactors? Linking personality type and anti-predator behaviour.

    Science.gov (United States)

    Jones, Katherine A; Godin, Jean-Guy J

    2010-02-22

    Response delays to predator attack may be adaptive, suggesting that latency to respond does not always reflect predator detection time, but can be a decision based on starvation-predation risk trade-offs. In birds, some anti-predator behaviours have been shown to be correlated with personality traits such as activity level and exploration. Here, we tested for a correlation between exploration behaviour and response latency time to a simulated fish predator attack in a fish species, juvenile convict cichlids (Amatitlania nigrofasciata). Individual focal fish were subjected to a standardized attack by a robotic fish predator while foraging, and separately given two repeated trials of exploration of a novel environment. We found a strong positive correlation between exploration and time taken to respond to the predator model. Fish that were fast to explore the novel environment were slower to respond to the predator. Our study therefore provides some of the first experimental evidence for a link between exploration behaviour and predator-escape behaviour. We suggest that different behavioural types may differ in how they partition their attention between foraging and anti-predator vigilance.

  9. Poole High Street study

    OpenAIRE

    Kilburn, David

    2007-01-01

    A presentation given to key decision makers within Poole to improve the retail offer in Poole High Street and leverage the benefit of improved town planning and the introduction of quality retail companies.

  10. Swimming Pool Safety

    Science.gov (United States)

    ... Prevention Listen Español Text Size Email Print Share Swimming Pool Safety Page Content ​What is the best way to keep my child safe around swimming pools? An adult should actively watch children at ...

  11. Swimming pool granuloma

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/001357.htm Swimming pool granuloma To use the sharing features on this page, please enable JavaScript. A swimming pool granuloma is a long-term (chronic) skin ...

  12. Validation of deterministic and Monte Carlo codes for neutronics calculation of the IRT-type research reactor

    Science.gov (United States)

    Shchurovskaya, M. V.; Alferov, V. P.; Geraskin, N. I.; Radaev, A. I.

    2017-01-01

    The results of the validation of a research reactor calculation using Monte Carlo and deterministic codes against experimental data and based on code-to-code comparison are presented. The continuous energy Monte Carlo code MCU-PTR and the nodal diffusion-based deterministic code TIGRIS were used for full 3-D calculation of the IRT MEPhI research reactor. The validation included the investigations for the reactor with existing high enriched uranium (HEU, 90 w/o) fuel and low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel.

  13. 野生金荞麦高黄酮含量愈伤组织中一个类异黄酮还原酶(IRL)基因的克隆与分析%Molecular cloning and characterization of a novel isoflavone reductase-like gene (FcIRL) from high flavonoids-producing callus of Fagopyrum cymosum

    Institute of Scientific and Technical Information of China (English)

    祝钦泷; 郭铁英; 眭顺照; 刘光德; 雷兴华; 罗莉莉; 李名扬

    2009-01-01

    Lignans are important defensive compounds in plants and have good biological activities protecting human health. In order to study the medicinal secondary metabolism of Fagopyrum cymosum (Trev.) Meisn, a traditional Chinese medicine with anti-tumor effect, a novel isoflavone reductase-like gene, FclRL, was cloned using RACE strategy from a cDNA library of high flavonoids-producing callus. The full-length eDNA of the FclRL was 1 217 bp (accession no. EU116032), which contained a 942 bp open reading frame (ORF) encoding a 313 amino acid protein. Two stop codons (TAG) and a putative polyadenylation signal ATAAA at 24 bp upstream from the polyadenylation site was found in 5' and 3' UTR, separately. And no intron was found in the genomic sequence yet. FclRL contained a predicted N-terminal acetylation site (M1-K5) and a NADPH-binding motif (G10-G-T-G13-Y-I-G16) in the N-terminal region, a conserved NmrA (nitrogen metabolite repression regulator) domain (V6-N244), multi-phosphorylation sites and one conserved N-glycosylation site (N214). Sequence homology comparison, phylogenetic analysis and advanced structures prediction all suggested that FclRL belonged to the class of pinoresinol-lariciresinol reductase (PLR), which is a key enzyme in synthetic pathway of 8-8'-linked lignans, with function in catalyzing reduction of pinoresinol and lariciresinol into secoisolariciresinol, and medicinal secondary metabolism and resistance in E cymosum.%木脂素是重要的植物防御物质之一,具有优良的生物学活性.为研究传统抗肿瘤良药野生金荞麦的药用次生代谢,用RACE与文库结合的方法,从野生金荞麦高黄酮含量愈伤系cDNA文库中快速克隆了一个类异黄酮还原酶(isoflavone reductase-like,IRL)基因FcIRL的全长eDNA(accession no.EUII6032).FcIRL cDNA全长1 217 bp,含有1个942 bp的开放阅读框(open reading frame,ORF),编码313个氨基酸,其5'LJTR区含有2个终止子TAG,3'UTR区含有推测的加尾信号ATAAA和polyA

  14. Accumulation of radioactive corrosion products on steel surfaces of VVER type nuclear reactors. I. 110mAg

    Science.gov (United States)

    Hirschberg, Gábor; Baradlai, Pál; Varga, Kálmán; Myburg, Gerrit; Schunk, János; Tilky, Péter; Stoddart, Paul

    Formation, presence and deposition of corrosion product radionuclides (such as 60Co, 51Cr, 54Mn, 59Fe and/or 110mAg) in the primary circuits of water-cooled nuclear reactors (PWRs) throw many obstacles in the way of normal operation. During the course of the work presented in this series, accumulations of such radionuclides have been studied at austenitic stainless steel type 08X18H10T (GOST 5632-61) surfaces (this austenitic stainless steel corresponds to AISI 321). Comparative experiments have been performed on magnetite-covered carbon steel (both materials are frequently used in some Soviet VVER type PWRs). For these laboratory-scale investigations a combination of the in situ radiotracer `thin gap' method and voltammetry is considered to be a powerful tool due to its high sensitivity towards the detection of the submonolayer coverages of corrosion product radionuclides. An independent technique (XPS) is also used to characterize the depth distribution and chemical state of various contaminants in the passive layer formed on austenitic stainless steel. In the first part of the series the accumulation of 110mAg has been investigated. Potential dependent sorption of Ag + ions (cementation) is found to be the predominant process on austenitic steel, while in the case of magnetite-covered carbon steel the silver species are mainly depleted in the form of Ag 2O. The XPS depth profile of Ag gives an evidence about the embedding of metallic silver into the entire passive layer of the austenitic stainless steel studied.

  15. The science of pooling

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, E.

    1995-10-01

    The pooling of data from radon studies is described. Pooling refers to the analysis of original data from several studies, not meta-analysis in which summary measures from published data are analyzed. A main objective for pooling is to reduce uncertainty and to obtain more precise estimates of risk than would be available from any single study.

  16. Electrical system regulations of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mello, Jose Roberto de; Madi Filho, Tufic, E-mail: jrmello@ipen.br, E-mail: tmfilho@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  17. The U.S. Geological Survey's TRIGA® reactor

    Science.gov (United States)

    DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

    2012-01-01

    The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

  18. Neutron beam facilities at the Australian Replacement Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, Shane; Robinson, Robert; Hunter, Brett [Physics Division, ANSTO, Lucas Heights (Australia)

    2001-03-01

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10{sup 14} n/cm{sup 2}/sec and a liquid D{sub 2} cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  19. Experimental determination of coolant flow pattern in hot and cold pools of PFBR using a large scale model

    Energy Technology Data Exchange (ETDEWEB)

    Indranil Banerjee; Rajesh, K.; AnandaRaj, M.; Venkata Ramanan, J.; Gopal, C.A.; Padmakumar, G.; Prakash, V.; Vaidyanathan, G. [Indira Gandhi Center for Atomic Research, Kalpakkam, 603102 (India)

    2005-07-01

    Full text of publication follows: The construction of Prototype Fast Breeder Reactor (PFBR) to generate 500 MWe has commenced at Kalpakkam, India. PFBR is a liquid sodium cooled pool type reactor with two secondary loops. The primary sodium pool is divided into hot pool and cold pool by means of Inner vessel. Cold sodium at 670 K is pumped through the core subassemblies and after absorbing the fission heat in the core, the sodium comes out and mixes with the hot pool at 820 K. This hot sodium exchanges heat with secondary sodium in Intermediate Heat Exchangers (IHX) which in turn transfers the heat to water in the steam generator leading to production of superheated steam to generate power. All the components like Control Plug (CP), IHX, Decay Heat Exchangers (DHX), Pump etc., are immersed in the primary sodium pool. The presence of these components influence the flow and velocity patterns of the coolant, in the hot and cold pools. The coolant behaviour in the pool is an indicator of the temperature pattern in the pool and the mechanical and thermal stresses induced on the immersed structures during transients is of significance for the safe operation of the reactor, designed for a life span of 40 years. Hence it is essential to understand the pattern of coolant flow and velocity patterns in hot and cold pools, particularly near IHX and Control plug. A 1:4 scale down model in stainless steel is constructed, simulating all the internal structures of the PFBR primary circuit for investigating the various parameters experimentally in water, to enhance the confidence in design of the primary system. The velocity distribution in the hot pool and cold pool at different regions, around the control plug, around the IHX inlet window were studied experimentally. As the coolant flow path is mainly influenced by the gravity force and inertia force, the study is conducted using Froude similitude. The magnitude of the velocity of the fluid at different points on the selected

  20. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  1. Genital mycotic infections with canagliflozin, a sodium glucose co-transporter 2 inhibitor, in patients with type 2 diabetes mellitus: a pooled analysis of clinical studies.

    Science.gov (United States)

    Nyirjesy, Paul; Sobel, Jack D; Fung, Albert; Mayer, Cristiana; Capuano, George; Ways, Kirk; Usiskin, Keith

    2014-06-01

    To characterize genital mycotic infections with canagliflozin, a sodium glucose co-transporter 2 inhibitor, in patients with type 2 diabetes mellitus (T2DM) using pooled data from Phase 3 studies. Genital mycotic infections with canagliflozin 100 and 300 mg were evaluated in Population 1 (N = 2313; mean exposure [weeks]: canagliflozin, 24.3; placebo, 23.8), including patients from four placebo-controlled studies, and Population 2 (N = 9439; mean exposure [weeks]: canagliflozin, 68.1; control, 64.4), including patients from eight placebo/active-controlled studies (including older patients and those with renal impairment or high cardiovascular disease risk). ClinicalTrials.gov, NCT01081834; NCT01106625; NCT01106677; NCT01106690; NCT01032629; NCT01064414; NCT01106651; NCT00968812. Adverse events suggestive of genital mycotic infections were recorded, with additional information collected using supplemental electronic case report forms. In Population 1, genital mycotic infection incidence was higher with canagliflozin 100 and 300 mg than placebo (95% confidence intervals excluded zero) in females (10.4%, 11.4%, 3.2%) and males (4.2%, 3.7%, 0.6%). These were generally mild to moderate in intensity, none were serious, and few led to discontinuation. Most events with canagliflozin were treated with antifungal therapies, and median symptom duration following treatment initiation was similar across groups; few patients had >1 event (females, 2.3%; males, 0.9%). Findings with canagliflozin 100 and 300 mg versus control were similar in Population 2 (females: 14.7%, 13.9%, 3.1%; males: 7.3%, 9.3%, 1.6%); a low proportion of males underwent circumcision across groups. Most events with canagliflozin occurred within the first 4 months in females and first year in males; no consistent evidence of dose dependence was observed. Key limitations included lack of laboratory confirmation for most events and variable treatment methods. Genital mycotic infection incidences

  2. Discrimination of source reactor type by multivariate statistical analysis of uranium and plutonium isotopic concentrations in unknown irradiated nuclear fuel material.

    Science.gov (United States)

    Robel, Martin; Kristo, Michael J

    2008-11-01

    The problem of identifying the provenance of unknown nuclear material in the environment by multivariate statistical analysis of its uranium and/or plutonium isotopic composition is considered. Such material can be introduced into the environment as a result of nuclear accidents, inadvertent processing losses, illegal dumping of waste, or deliberate trafficking in nuclear materials. Various combinations of reactor type and fuel composition were analyzed using Principal Components Analysis (PCA) and Partial Least Squares Discriminant Analysis (PLSDA) of the concentrations of nine U and Pu isotopes in fuel as a function of burnup. Real-world variation in the concentrations of (234)U and (236)U in the fresh (unirradiated) fuel was incorporated. The U and Pu were also analyzed separately, with results that suggest that, even after reprocessing or environmental fractionation, Pu isotopes can be used to determine both the source reactor type and the initial fuel composition with good discrimination.

  3. A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Antonio Carlos Pinto Dias

    1993-12-31

    The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

  4. Hybrid reactors. [Fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  5. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

    OpenAIRE

    2014-01-01

    To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests...

  6. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  7. NSGA-II Algorithm with a Local Search Strategy for Multiobjective Optimal Design of Dry-Type Air-Core Reactor

    Directory of Open Access Journals (Sweden)

    Chengfen Zhang

    2015-01-01

    Full Text Available Dry-type air-core reactor is now widely applied in electrical power distribution systems, for which the optimization design is a crucial issue. In the optimization design problem of dry-type air-core reactor, the objectives of minimizing the production cost and minimizing the operation cost are both important. In this paper, a multiobjective optimal model is established considering simultaneously the two objectives of minimizing the production cost and minimizing the operation cost. To solve the multi-objective optimization problem, a memetic evolutionary algorithm is proposed, which combines elitist nondominated sorting genetic algorithm version II (NSGA-II with a local search strategy based on the covariance matrix adaptation evolution strategy (CMA-ES. NSGA-II can provide decision maker with flexible choices among the different trade-off solutions, while the local-search strategy, which is applied to nondominated individuals randomly selected from the current population in a given generation and quantity, can accelerate the convergence speed. Furthermore, another modification is that an external archive is set in the proposed algorithm for increasing the evolutionary efficiency. The proposed algorithm is tested on a dry-type air-core reactor made of rectangular cross-section litz-wire. Simulation results show that the proposed algorithm has high efficiency and it converges to a better Pareto front.

  8. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2; Analisis conceptual de un modelo preliminar para el estudio de la inestabilidad en la operacion normal de un reactor de circulacion natural tipo ESBWR, usando Relap5/Mod 3.2

    Energy Technology Data Exchange (ETDEWEB)

    Ojeda S, J.; Morales S, J.; Chavez M, C. [UNAM, Facultad de Ingenieria, Circuito Exterior, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: j.os.ojeda@hotmail.com

    2009-10-15

    This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

  9. Pressurized Water Reactor Nuclear Power Plant Spent Fuel Pool Spray System Design%压水堆核电站乏燃料池喷淋系统设计

    Institute of Scientific and Technical Information of China (English)

    苏夏

    2013-01-01

      第三代非能动压水堆核电站AP1000中首次为乏燃料池设置了喷淋系统,在超设计基准事故或恐怖袭击导致乏燃料池水排空时,为乏燃料提供冷却。喷淋系统设计中的两个重要指标是喷淋覆盖面积和单位面积有效喷淋流量。设计者应基于喷嘴性能试验结果,根据乏燃料池结构尺寸和乏燃料特性,确定喷淋流量、喷嘴数量和布置方式等参数,完成系统设计,提供足够冷却流量。%  Spray system of spent fuel pool is first designed in AP1000, it can provide spray water to cool the spent fuel in a beyond design basis event or a terror attack that drains the pool. The two most important factors of spray system are the coverage pattern and the effective flow density. The spray flowrate, the nozzle number and their location should be designed based on the spray nozzle test results, the spent fuel pool structure and the spent fuel character to achieve the intent of providing enough cooling.

  10. Improving the performance of the Egyptian second testing nuclear research reactor using interval type-2 fuzzy logic controller tuned by modified biogeography-based optimization

    Energy Technology Data Exchange (ETDEWEB)

    Sayed, M.M., E-mail: M.M.Sayed@ieee.org; Saad, M.S.; Emara, H.M.; Abou El-Zahab, E.E.

    2013-09-15

    Highlights: • A modified version of the BBO was proposed. • A novel method for interval type-2 FLC design tuned by MBBO was proposed. • The performance of the ETRR-2 was improved by using IT2FLC tuned by MBBO. -- Abstract: Power stabilization is a critical issue in nuclear reactors. The conventional proportional derivative (PD) controller is currently used in the Egyptian second testing research reactor (ETRR-2). In this paper, we propose a modified biogeography-based optimization (MBBO) algorithm to design the interval type-2 fuzzy logic controller (IT2FLC) to improve the performance of the Egyptian second testing research reactor (ETRR-2). Biogeography-based optimization (BBO) is a novel evolutionary algorithm that is based on the mathematical models of biogeography. Biogeography is the study of the geographical distribution of biological organisms. In the BBO model, problem solutions are represented as islands, and the sharing of features between solutions is represented as immigration and emigration between the islands. A modified version of the BBO is applied to design the IT2FLC to get the optimal parameters of the membership functions of the controller. We test the optimal IT2FLC obtained by modified biogeography-based optimization (MBBO) using the integral square error (ISE) and is compared with the currently used PD controller.

  11. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  12. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-07-15

    Due to the corrosion problems of the aluminum coating of the reactor pool, a periodic inspections program by ultrasound to evaluate the advance grade and the corrosion speed was settled down. This inspections have shown the necessity to repair some areas, in those that the slimming is significant, of not making it can arrive to the water escape of the reactor pool. The objective of the repair is to place patches of plates of 1/4 inch aluminum thickness in the areas of the reactor 'liner', in those that it has been detected by ultrasound a smaller thickness or similar to 3 mm. To carry out this the fuels are move (of the core and those that are decaying) to a temporary storage, the structure of the core is confined in a tank that this placed inside the pool of the reactor, a shield is placed in the thermal column and it is completely extracted the water for to leave uncover the 'liner' of the reactor. (Author)

  13. Start-up and steady-state conditions of an Anaerobic Hybrid Reactor (AHR) using mini-filters composed with two types of support medium operating under low loading rates

    National Research Council Canada - National Science Library

    Silva, Vivian Galdino da; Campos, Cláudio Milton Montenegro; Pereira, Erlon Lopes; Silva, Júlia Ferreira da

    2011-01-01

    ...) removing organic matter of coffee wastewater with low concentration. The AHR was built similar to an UASB reactor, however the interior was filled with mini-filters composed by two types of support materials...

  14. The rehabilitation/upgrading of Philippine Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Renato, T. Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

    1998-10-01

    The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

  15. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  16. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  17. Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

    2004-07-01

    The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

  18. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  19. A modular method for the extraction of DNA and RNA, and the separation of DNA pools from diverse environmental sample types

    DEFF Research Database (Denmark)

    Lever, Mark; Torti, Andrea; Eickenbusch, Philip

    2015-01-01

    tests, in which permutations of all nucleic acid extraction steps were compared. The final modular protocol is suitable for extractions from igneous rock, air, water, and sediments. Sediments range from high-biomass, organic rich coastal samples to samples from the most oligotrophic region of the world......A method for the extraction of nucleic acids from a wide range of environmental samples was developed. This method consists of several modules, which can be individually modified to maximize yields in extractions of DNA and RNA or separations of DNA pools. Modules were designed based on elaborate...

  20. Strength training increases the size of the satellite cell pool in type I and II fibres of chronically painful trapezius muscle in females

    DEFF Research Database (Denmark)

    Mackey, Abigail; Andersen, Lars L; Frandsen, Ulrik

    2011-01-01

    While strength training has been shown to be effective in mediating hypertrophy and reducing pain in trapezius myalgia, responses at the cellular level have not previously been studied. This study investigated the potential of strength training targeting the affected muscles (SST, n = 18......) and general fitness training (GFT, n = 16) to augment the satellite cell (SC) and macrophage pools in the trapezius muscles of women diagnosed with trapezius myalgia. A group receiving general health information (REF, n = 8) served as a control. Muscle biopsies were collected from the trapezius muscles...... hypertrophy (r = -0.669, P = 0.005). SST also resulted in a 74% enhancement of the trapezius macrophage content (P

  1. Jordan's First Research Reactor Project: Driving Forces, Present Status and the Way Ahead

    Energy Technology Data Exchange (ETDEWEB)

    Xoubi, Ned, E-mail: Ned@Xoubi.co [Jordan Atomic Energy Commission (JAEC), P.O.Box 70, Shafa Badran, 11934 Amman (Jordan)

    2011-07-01

    In a gigantic step towards establishing Jordan's nuclear power program, Jordan Atomic Energy Commission (JAEC) is building the first nuclear research and test reactor in the Kingdom. The new reactor will serve as the focal point for Jordan Center for Nuclear Research (JCNR), a comprehensive state of the art nuclear center not only for Jordan but for the whole region, the center will include in addition to the reactor a radioisotopes production plant, a nuclear fuel fabrication plant, a cold neutron source (CNS), a radioactive waste treatment facility, and education and training center. The JRTR reactor is the only research reactor new build worldwide in 2010, it is a 5 MW light water open pool multipurpose reactor, The reactor core is composed of 18 fuel assemblies, MTR plate type, with 19.75% enriched uranium silicide (U{sub 3}Si{sub 2}) in an aluminum matrix. It is reflected on all sides by beryllium and graphite blocks. Reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45x10{sup 14} cm{sup -2}s{sup -1}. The reactor reactivity is controlled by four Hafnium Control Absorber Rods (CAR). Jordan Center for Nuclear Research is located in Ramtha city, it is owned by Jordan Atomic Energy Commission (JAEC), and is contracted to Korea Atomic Energy Research Institute (KAERI) and Daewoo E and C. The JCNR project is a 56 months EPC fixed price contract for the design engineering, construction, and commissioning the JCNR reactor, and other nuclear facilities. The project presents many challenges for both the owner and the contractor, being the first nuclear reactor for Jordan, and the first nuclear export for Korea. The driving forces, present status and the way ahead will be presented in this paper. (author)

  2. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  3. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  4. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  5. China experimental fast reactor; Le reacteur rapide experimental chinois

    Energy Technology Data Exchange (ETDEWEB)

    Tianmin, X. [Institut d' Ingenierie Nucleaire de Pekin (China); Cunren, L. [Centre d' Etude de Surete de Pekin (China)

    2007-07-15

    The Chinese experimental fast reactor (CEFR) is a pool-type sodium-cooled fast reactor whose short term purposes are: -) the validation of computer codes, -) the check of the relevance of standards, and -) the gathering of experimental data on fast reactors. On the long term the expectations will focus on: -) gaining experience in fast reactor operations, -) the testing of nuclear fuels and materials, and -) the study of sodium compounds. The main technical features of CEFR are: -) thermal power output: 65 MW (electrical power output: 20 MW), -) size of the core: height: 45 cm, diameter: 60 cm, -) maximal linear output: 430 W/cm, -) neutron flux: 3.7*10{sup 15} n/cm{sup 2}/s, -) input/output sodium temperature: 360 / 530 Celsius degrees, -) 2 loops for the primary system and 2 loops for the secondary system. The temperature coefficient and the power coefficient are settled to stay negative for any change in the values of the core parameters. The installation of the reactor vessel will be completed by mid 2007. The first criticality of CEFR is expected during the first semester of 2010. (A.C.)

  6. A Study on the Electromagnet Thrust force Characteristics of Newly Proposed Hybrid Bottom-mounted Control Rod Drive Mechanism for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Cho, Yeong Garp; Choi, Myoung Hwan; Yu, Je Yong; Kim, Ji Ho; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The control rod drive mechanism (CRDM) is the part of reactor regulating system (RRS), which is located in the reactor pool top (Top-mounted) or the room below the reactor pool (Bottom-mounted). The function of the CRDM is to insert, withdraw or maintain neutron absorbing material at any required position within the reactor core, in order to the reactivity control of the core. There are so many kinds of CRDM, such as magneticjack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we have investigated the movable coil electromagnetic drive mechanism (MCEDM) which is new scheme for the reactor control rod adopted by China Advanced Research Reactor (CARR) as shown in Fig.1. To improve a better function of the electromagnetic and magnetic characteristics, new model CRDM, which is named a hybrid bottommounted CRDM (HBCRDM), is proposed. Especially in order to achieve improved thrust force, numerical magnetic field calculations between MCEDM and HBCRDM have been carried out and the HBCRDM FEM results have been compared with the MCEDM FEM results, and FEM results are summarized in the following sections

  7. Análisis para la modelación y optimización geométrica de un reactor tipo tornillo sin-fin empleando el método de grafos dicromáticos//Analysis for geometric modeling and optimization of a worm type reactor using the method of dichromatic graph

    Directory of Open Access Journals (Sweden)

    Armando Díaz-Concepción

    2015-09-01

    Full Text Available En el presente trabajo se realiza la modelación, simulación y optimización de un reactor utilizado en las plantas para la obtención de un alimento animal, sobre la base de la predigestión del bagacillo de caña y el hidróxido de calcio en presencia de vapor denominado PREDICAL utilizando grafos dicromáticos. Se obtuvo el modelo matemático para el diseño del reactor, donde se vinculan las variables geométricas y tecnológicas. El modelo formulado permitió la optimización de la variable costo a partir de minimizar la variable geométrica diámetro exterior del reactor. Palabras claves: modelación reactor tipo tornillo sinfin, grafos dicromáticos, modelo matemático________________________________________________________________________________AbstractThe present work performs modeling, simulation and optimization of a reactor used in plants for the obtencion of animal feed. It's made on the basis of pre-digestion of cane bagasse and calcium hydroxide in the presence of steam called PREDICAL and using dichromatic graphs. It was achieved the mathematical model for the design of the reactor, where are linked geometric and technological variables. The model developed allowed cost optimization based on minimize the geometric variable outside diameter of the reactor. Key words: worm type reactor modeling, dichromatic graphs, mathematical model.

  8. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J.

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  9. Biofiltration of a styrene/acetone vapor mixture in two reactor types under conditions of styrene overloading

    Directory of Open Access Journals (Sweden)

    Lubos Zapotocky

    2014-10-01

    Full Text Available This aim of study was to compare the performance of a biofilter (BF and trickle bed reactor (TBR under increased styrene loading with a constant acetone load, 2 gc/m3/h. At styrene loading rates up to 30 gc/m3/h, the BF showed higher styrene removal than TBR. However, the BF efficiency started to drop beyond this threshold loading and could never reach steady state, whereas the TBR continued to yield a 50% styrene removal. The acetone removal remained constant (93-98% in both the reactors at any styrene loading. Once the overloading was lifted, the BF recovered within 26 min, whereas the TBR efficiency bounced back only to 95%, gradually returning to complete removal only in 10 h.

  10. Influence of the type and source of inoculum on the start-up of anammox sequencing batch reactors (SBRs).

    Science.gov (United States)

    Guerrero, Lorna; Van Diest, Federico; Barahona, Andrea; Montalvo, Silvio; Borja, Rafael

    2013-01-01

    Anammox (anaerobic ammonium oxidation) is an attractive option for the treatment of wastewaters with a low carbon/nitrogen ratio. This is due to its low operating costs when compared to the classical nitrification-denitrification processes. However, one of the main disadvantages of the Anammox process is slow biomass growth, meaning a relatively slow reactor start-up. This becomes even more complicated when Anammox microorganisms are not present in the inoculum. Four inocula were studied for the start-up of Anammox sequencing batch reactors (SBRs) 2 L in volume agitated at 100 rpm, one of them using zeolite as a microbial support. Two inocula were taken from UASB reactors and two from aerobic reactors (activated sludge and SBR). The Anammox SBRs studied were operated at 36 ± 0.5°C. The results showed that the only inoculum that enabled the enrichment of the Anammox biomass came from an activated sludge plant treating wastewaters from a poultry slaughterhouse. This plant was designed for organic matter degradation and nitrogen removal (nitrification). This could explain the presence of Anammox microorganisms. This SBR operated without zeolite and achieved nitrite and ammonium removals of 96.3% and 68.4% respectively, at a nitrogen loading rate (NLR) of 0.1 kg N/m(3)/d in both cases. The lower ammonium removal was due to the fact that a sub-stoichiometric amount of nitrite (1 molar ratio) was fed. The specific Anammox activity (SAA) achieved was 0.18 g N/g VSS/d.

  11. MCTP, a code for the thermo-mechanical analysis of a fuel rod of BWR type reactors (Neutron part); MCTP, un codigo para el analisis termo-mecanico de una barra combustible de reactores tipo BWR (Parte Neutronica)

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez L, H.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: hhl@nuclear.inin.mx

    2003-07-01

    In the National Institute of Nuclear Research of Mexico a code for the thermo-mechanical analysis of the fuel rods of the BWR type reactors of the Nucleo electric Central of Laguna Verde is developed. The code solves the diffusion equation in cylindrical coordinates with several energy groups. The code, likewise, calculates the temperature distribution and power distribution in those fuel rods. The code is denominated Multi groups With Temperatures and Power (MCTP). In the code, the energy with which the fission neutrons are emitted it is divided in six groups. They are also considered the produced perturbations by the changes in the temperatures of the materials that constitute the fuel rods, the content of fission products, the uranium consumption and in its case the gadolinium, as well as the plutonium production. In this work there are present preliminary results obtained with the code, using data of operation of the Nucleo electric Central of Laguna Verde. (Author)

  12. Method for calculating coolant resonance frequencies under normal and accident conditions in nuclear power plants with WWER-type pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N. (Moskovskij Ehnergeticheskij Inst. (USSR))

    1983-03-01

    Mathematical models are proposed for calculating acoustic oscillation resonance frequencies in the coolant in various components of the WWER type primary circuit (core, steam generator, pressurizer, piping). Due to the correspondence between model calculations and experimental results obtained in operating nuclear power plants, the developed models can be used for practical calculations. The possibility of calculating the eigenfrequencies of the coolant oscillation under different operating conditions leads to the interpretation of operational data, to the analysis of operational conditions, to the detection of coolant boiling in the reactor, and to design changes in order to prevent resonance oscillations within the coolant.

  13. PDA: Pooled DNA analyzer

    Directory of Open Access Journals (Sweden)

    Lin Chin-Yu

    2006-04-01

    Full Text Available Abstract Background Association mapping using abundant single nucleotide polymorphisms is a powerful tool for identifying disease susceptibility genes for complex traits and exploring possible genetic diversity. Genotyping large numbers of SNPs individually is performed routinely but is cost prohibitive for large-scale genetic studies. DNA pooling is a reliable and cost-saving alternative genotyping method. However, no software has been developed for complete pooled-DNA analyses, including data standardization, allele frequency estimation, and single/multipoint DNA pooling association tests. This motivated the development of the software, 'PDA' (Pooled DNA Analyzer, to analyze pooled DNA data. Results We develop the software, PDA, for the analysis of pooled-DNA data. PDA is originally implemented with the MATLAB® language, but it can also be executed on a Windows system without installing the MATLAB®. PDA provides estimates of the coefficient of preferential amplification and allele frequency. PDA considers an extended single-point association test, which can compare allele frequencies between two DNA pools constructed under different experimental conditions. Moreover, PDA also provides novel chromosome-wide multipoint association tests based on p-value combinations and a sliding-window concept. This new multipoint testing procedure overcomes a computational bottleneck of conventional haplotype-oriented multipoint methods in DNA pooling analyses and can handle data sets having a large pool size and/or large numbers of polymorphic markers. All of the PDA functions are illustrated in the four bona fide examples. Conclusion PDA is simple to operate and does not require that users have a strong statistical background. The software is available at http://www.ibms.sinica.edu.tw/%7Ecsjfann/first%20flow/pda.htm.

  14. A simplified model of decontamination by BWR steam suppression pools

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A.

    1997-05-01

    Phenomena that can decontaminate aerosol-laden gases sparging through steam suppression pools of boiling water reactors during reactor accidents are described. Uncertainties in aerosol properties, aerosol behavior within gas bubbles, and bubble behavior in plumes affect predictions of decontamination by steam suppression pools. Uncertainties in the boundary and initial conditions that are dictated by the progression of severe reactor accidents and that will affect predictions of decontamination by steam suppression pools are discussed. Ten parameters that characterize boundary and initial condition uncertainties, nine parameters that characterize aerosol property and behavior uncertainties, and eleven parameters that characterize uncertainties in the behavior of bubbles in steam suppression pools are identified. Ranges for the values of these parameters and subjective probability distributions for parametric values within the ranges are defined. These uncertain parameters are used in Monte Carlo uncertainty analyses to develop uncertainty distributions for the decontamination that can be achieved by steam suppression pools and the size distribution of aerosols that do emerge from such pools. A simplified model of decontamination by steam suppression pools is developed by correlating features of the uncertainty distributions for total decontamination factor, DF(total), mean size of emerging aerosol particles, d{sub p}, and the standard deviation of the emerging aerosol size distribution, {sigma}, with pool depth, H. Correlations of the median values of the uncertainty distributions are suggested as the best estimate of decontamination by suppression pools. Correlations of the 10 percentile and 90 percentile values of the uncertainty distributions characterize the uncertainty in the best estimates. 295 refs., 121 figs., 113 tabs.

  15. Possible pressurized thermal shock events during large primary to secondary leakage. The Hungarian AGNES project and PRISE accident scenarios in VVER-440/V213 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Perneczky, L. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

    1997-12-31

    Nuclear power plants of WWER-440/213-type have several special features. Consequently, the transient behaviour of such a reactor system should be different from the behaviour of the PWRs of western design. The opening of the steam generator (SG) collector cover, as a specific primary to secondary circuit leakage (PRISE) occurring in WWER-type reactors happened first time in Rovno NPP Unit I on January 22, 1982. Similar accident was studied in the framework of IAEA project RER/9/004 in 1987-88 using the RELAP4/mod6 code. The Hungarian AGNES (Advanced General and New Evaluation of Safety) project was performed in the period 1991-94 with the aim to reassess the safety of the Paks NPP using state-of-the-art techniques. The project comprised three type of analyses for the primary to secondary circuit leakages: Design Basis Accident (DBA) analyses, Pressurized Thermal Shock (PTS) study and deterministic analyses for Probabilistic Safety Analysis (PSA). Major part of the thermohydraulic analyses has been performed by the RELAP5/mod2.5/V251 code version with two input models. 32 refs.

  16. Mathematical Modeling for Simulation of Nuclear Reactor Analysis

    OpenAIRE

    Salah Ud-Din Khan; Shahab Ud-Din Khan

    2013-01-01

    In this paper, we have developed a mathematical model for the nuclear reactor analysis to be implemented in the nuclear reactor code. THEATRe is nuclear reactor analysis code which can only work for the cylindrical type fuel reactor and cannot applicable for the plate type fuel nuclear reactor. Therefore, the current studies encompasses on the modification of THEATRe code for the plate type fuel element. This mathematical model is applicable to the thermal analysis of the reactor which is ver...

  17. Life time of nuclear power plants and new types of reactors; La duree de vie des centrales nucleaires et les nouveaux types de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  18. Vitamin D Pooling Project

    Science.gov (United States)

    The Vitamin D Pooling Project of Rarer Cancers brought together investigators from 10 cohorts to conduct a large prospective epidemiologic study of the association between vitamin D status and seven rarer cancers.

  19. Verification Calculation Results to Validate the Procedures and Codes for Pin-by-Pin Power Computation in VVER Type Reactors with MOX Fuel Loading

    Energy Technology Data Exchange (ETDEWEB)

    Chizhikova, Z.N.; Kalashnikov, A.G.; Kapranova, E.N.; Korobitsyn, V.E.; Manturov, G.N.; Tsiboulia, A.A.

    1998-12-01

    One of the important problems for ensuring the VVER type reactor safety when the reactor is partially loaded with MOX fuel is the choice of appropriate physical zoning to achieve the maximum flattening of pin-by-pin power distribution. When uranium fuel is replaced by MOX one provided that the reactivity due to fuel assemblies is kept constant, the fuel enrichment slightly decreases. However, the average neutron spectrum fission microscopic cross-section for {sup 239}Pu is approximately twice that for {sup 235}U. Therefore power peaks occur in the peripheral fuel assemblies containing MOX fuel which are aggravated by the interassembly water. Physical zoning has to be applied to flatten the power peaks in fuel assemblies containing MOX fuel. Moreover, physical zoning cannot be confined to one row of fuel elements as is the case with a uniform lattice of uranium fuel assemblies. Both the water gap and the jump in neutron absorption macroscopic cross-sections which occurs at the interface of fuel assemblies with different fuels make the problem of calculating space-energy neutron flux distribution more complicated since it increases nondiffusibility effects. To solve this problem it is necessary to update the current codes, to develop new codes and to verify all the codes including nuclear-physical constants libraries employed. In so doing it is important to develop and validate codes of different levels--from design codes to benchmark ones. This paper presents the results of the burnup calculation for a multiassembly structure, consisting of MOX fuel assemblies surrounded by uranium dioxide fuel assemblies. The structure concerned can be assumed to model a fuel assembly lattice symmetry element of the VVER-1000 type reactor in which 1/4 of all fuel assemblies contains MOX fuel.

  20. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  1. State of art report for critical flow model to analyze a break flow in pressurizer of integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yeon Moon; Lee, D. J.; Yoon, J. H.; Kim, J. P.; Kim, H. Y

    1999-03-01

    At a critical flow condition, the flow rate can't exceed a maximum value for given upstream conditions and the limited flow rate is called as a critical flow rate. The phenomena of critical flow occur at the discharge of a single phase gas or subcooled water through nozzles and pipes. Among the previous researches on critical flow, many accurate correlations on pressure, temperature and flow rate are represented for the single phase gas. However, for the two phase critical flow, the results of previous work showed that there was a large discrepancy between the analytical and experimental data and the data were in agreement for the limited thermodynamic conditions. Thus, further studies are required to enhance the two phase critical flow model. In the integral reactor, the critical flows of nitrogen gas and subcooled water are expected for the break of gas cylinder pipeline connected to the pressurizer. It requires that the inlet shape of the pipe and the nitrogen gas effect should be considered for the critical flow of integral reactor. The nitrogen gas exist in the pressurizer may affect the flow rate of primary coolant, which has been considered only for a few previous researches. Thus, the evaluation of the effect of the nitrogen on the critical flow gas should be preceded for the proper analysis of the critical flow in the integral reactor. In this report, not only the essences of previous work on critical flow were investigated and summarized but also the effect of nitrogen gas and the inlet shape of the pipe on the critical flow were also investigated. (author)

  2. State of art report for critical flow model to analyze a break flow in pressurizer of integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yeon Moon; Lee, D. J.; Yoon, J. H.; Kim, J. P.; Kim, H. Y

    1999-03-01

    At a critical flow condition, the flow rate can't exceed a maximum value for given upstream conditions and the limited flow rate is called as a critical flow rate. The phenomena of critical flow occur at the discharge of a single phase gas or subcooled water through nozzles and pipes. Among the previous researches on critical flow, many accurate correlations on pressure, temperature and flow rate are represented for the single phase gas. However, for the two phase critical flow, the results of previous work showed that there was a large discrepancy between the analytical and experimental data and the data were in agreement for the limited thermodynamic conditions. Thus, further studies are required to enhance the two phase critical flow model. In the integral reactor, the critical flows of nitrogen gas and subcooled water are expected for the break of gas cylinder pipeline connected to the pressurizer. It requires that the inlet shape of the pipe and the nitrogen gas effect should be considered for the critical flow of integral reactor. The nitrogen gas exist in the pressurizer may affect the flow rate of primary coolant, which has been considered only for a few previous researches. Thus, the evaluation of the effect of the nitrogen on the critical flow gas should be preceded for the proper analysis of the critical flow in the integral reactor. In this report, not only the essences of previous work on critical flow were investigated and summarized but also the effect of nitrogen gas and the inlet shape of the pipe on the critical flow were also investigated. (author)

  3. Assessment of the thorium fuel cycle in power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled.

  4. Covalent immobilization of catalase onto spacer-arm attached modified florisil: characterization and application to batch and plug-flow type reactor systems.

    Science.gov (United States)

    Alptekin, Ozlem; Tükel, S Seyhan; Yildirim, Deniz; Alagöz, Dilek

    2011-12-10

    Catalase was covalently immobilized onto florisil via glutaraldehyde (GA) and glutaraldehyde+6-amino hexanoic acid (6-AHA) (as a spacer arm). Immobilizations of catalase onto modified supports were optimized to improve the efficiency of the overall immobilization procedures. The V(max) values of catalase immobilized via glutaraldehyde (CIG) and catalase immobilized via glutaraldehyde+6-amino hexanoic acid (CIG-6-AHA) were about 0.6 and 3.4% of free catalase, respectively. The usage of 6-AHA as a spacer arm caused about 40 folds increase in catalytic efficiency of CIG-6-AHA (8.3 × 10⁵ M⁻¹ s⁻¹) as compared to that of CIG (2.1 × 10⁴ M⁻¹ s⁻¹). CIG and CIG-6-AHA retained 67 and 35% of their initial activities at 5 °C and 71 and 18% of their initial activities, respectively at room temperature at the end of 6 days. Operational stabilities of CIG and CIG-6-AHA were investigated in batch and plug-flow type reactors. The highest total amount of decomposed hydrogen peroxide (TAD-H₂O₂) was determined as 219.5 μmol for CIG-6-AHA in plug-flow type reactor. Copyright © 2011 Elsevier Inc. All rights reserved.

  5. Experimental results of acetone hydrogenation on a heat exchanger type reactor for solar chemical heat pump; Solar chemical heat pump ni okeru acetone suisoka hanno netsu kaishu jikken

    Energy Technology Data Exchange (ETDEWEB)

    Takashima, T.; Doi, T.; Tanaka, T.; Ando, Y. [Electrotechnical Laboratory, Tsukuba (Japan); Miyahara, R.; Kamoshida, J. [Shibaura Institute of Technology, Tokyo (Japan)

    1996-10-27

    With the purpose of converting solar heat energy to industrial heat energy, an experiment of acetone hydrogenation was carried out using a heat exchanger type reactor that recovers heat generated by acetone hydrogenation, an exothermic reaction, and supplies it to an outside load. In the experiment, a pellet-like activated carbon-supported ruthenium catalyst was used for the acetone hydrogenation with hydrogen and acetone supplied to the catalyst layer at a space velocity of 400-1,200 or so. In the external pipe of the double-pipe type reactor, a heating medium oil was circulated in parallel with the flow of the reactant, with the heat of reaction recovered that was generated from the acetone hydrogenation. In this experiment, an 1wt%Ru/C catalyst and a 5wt%Ru/C catalyst were used so as to examine the effects of variation in the space velocity. As a result, from the viewpoint of recovering the heat of reaction, it was found desirable to increase the reaction speed by raising catalytic density and also to supply the reactant downstream inside the reaction pipe by increasing the space velocity. 1 ref., 6 figs., 1 tab.

  6. Development of two-dimensional hot pool model

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Hahn, H. D

    2000-05-01

    During a normal reactor scram, the heat generation is reduced almost instantaneously while the coolant flow rate follows the pump coast-down. This mismatch between power and flow results in a situation where the core flow entering the hot pool is at a lower temperature than the temperature of the bulk pool sodium. This temperature difference leads to thermal stratification. Thermal stratification can occur in the hot pool region if the entering coolant is colder than the existing hot pool coolant and the flow momentum is not large enough to overcome the negative buoyancy force. Since the fluid of hot pool enters IHX{sub s}, the temperature distribution of hot pool can alter the overall system response. Hence, it is necessary to predict the pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHX{sub s} and its contribution to the net buoyancy head. Therefore, in this study two-dimensional hot pool model is developed instead of existing one-dimensional model to predict the hot pool coolant temperature and velocity distribution more accurately and is applied to the SSC-K code.

  7. Experiments and Characterization of the Two-Phase Flow Driven Particulate Debris Spreading in the Pool

    OpenAIRE

    Konovalenko, Alexander; Basso, Simone; Kudinov, Pavel

    2014-01-01

    Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. Success of the strategy is contingent upon effectiveness of natural circulation in removing the decay heat generated by the porous debris bed. Geometrical configuration of the bed is one of the factors which affect coolability of the bed. Boiling and two-phase turbulent flows in the pool serve as ...

  8. Reverse-Bumpy-Ball-Type-Nanoreactor-Loaded Nylon Membranes as Peroxidase-Mimic Membrane Reactors for a Colorimetric Assay for H₂O₂.

    Science.gov (United States)

    Tong, Ying; Jiao, Xiangyu; Yang, Hankun; Wen, Yongqiang; Su, Lei; Zhang, Xueji

    2016-04-01

    Herein we report for the first time fabrication of reverse bumpy ball (RBB)-type-nanoreactor-based flexible peroxidase-mimic membrane reactors (MRs). The RBB-type nanoreactors with gold nanoparticles embedded in the inner walls of carbon shells were loaded on nylon membranes through a facile filtration approach. The as-prepared flexible catalytic membrane was studied as a peroxidase-mimic MR. It was found that the obtained peroxidase-mimic MR could exhibit several advantages over natural enzymes, such as facile and good recyclability, long-term stability and easy storage. Moreover, the RBB NS-modified nylon MRs as a peroxidase mimic provide a useful colorimetric assay for H₂O₂.

  9. Reverse-Bumpy-Ball-Type-Nanoreactor-Loaded Nylon Membranes as Peroxidase-Mimic Membrane Reactors for a Colorimetric Assay for H2O2

    Science.gov (United States)

    Tong, Ying; Jiao, Xiangyu; Yang, Hankun; Wen, Yongqiang; Su, Lei; Zhang, Xueji

    2016-01-01

    Herein we report for the first time fabrication of reverse bumpy ball (RBB)-type-nanoreactor-based flexible peroxidase-mimic membrane reactors (MRs). The RBB-type nanoreactors with gold nanoparticles embedded in the inner walls of carbon shells were loaded on nylon membranes through a facile filtration approach. The as-prepared flexible catalytic membrane was studied as a peroxidase-mimic MR. It was found that the obtained peroxidase-mimic MR could exhibit several advantages over natural enzymes, such as facile and good recyclability, long-term stability and easy storage. Moreover, the RBB NS-modified nylon MRs as a peroxidase mimic provide a useful colorimetric assay for H2O2. PMID:27043575

  10. Experimental Investigation and Analysis on Temperature Distribution in Water Pool of External Reactor Vessel Cooling System%压力容器外部冷却系统冷却水池温度场的实验研究及分析

    Institute of Scientific and Technical Information of China (English)

    李永春; 李飞; 程旭; 杨燕华

    2013-01-01

    压力容器外部非能动冷却系统采用换料水池作为冷却水源。在浮升力驱动的自然循环流动作用下,冷却水池内会逐渐出现热分层现象。本实验基于先进压水堆压力容器外部冷却系统模拟装置REPEC实验回路,通过测量实验系统内冷却水箱的温度场空间分布,对冷却水池的热分层与混合现象、发展规律和主要影响因素进行了实验分析。结果表明:实验水箱内温度场分布差异主要表现在高度方向;循环流量是影响热分层的重要参数,而水箱工质初始温度的影响非常微弱;针对本实验的无量纲一维瞬态温度场方程分析表明,水箱内温度场的发展规律主要受对流传热控制。%In the external reactor vessel passive cooling system ,the refueling water pool is used as the heat trap .Driven by buoyancy force ,thermal stratification phenomena occur spontaneously in the pool . T he thermal stratification and mixing phenomena , development process and main influencing factors in the refueling water pool with the REPEC test facility were studied .The experimental results show that the temperature distribution in the experimental water tank varies with height ,the circulation flow is one of the most important factors w hich influence the thermal stratification process in the water tank ,and the effect of initial water temperature is negligible .The analysis result of the 1-D non-dimensional transient temperature field equation show s that the development of the temperature field in the water tank is mainly controlled by convective heat transfer .

  11. The importance of carry out studies about the use of passive autocatalytic recombiners for hydrogen control in reactors type ESBWR; La importancia de realizar estudios sobre el uso de recombinadores autocataliticos pasivos para control de hidrogeno en reactores tipo ESBWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J.; Morales S, J. B. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)], e-mail: jersonsanchez@gmail.com

    2009-10-15

    A way to satisfy and to guarantee the energy necessities in the future is increasing in a gradual way the creation of nuclear power plants, introducing advanced designs in its systems that contribute in way substantial in the security of the same nuclear plants. The tendency of new designs of these nuclear plants is the incorporation of systems more reliable and sure, and that the operation does not depend on external factors as the electric power, motors diesel or the action of the operator of nuclear plant, what is known as security passive systems. In this sense, the passive autocatalytic recombiners are a contribution toward the use of this type of systems. At the present time it is had studies of the incorporation of passive autocatalytic recombiners in nuclear plants in operation and that they have contributed to minimize the danger associated to hydrogen. The present work contains a first approach to the study of hydrogen recombiners incorporation in advanced nuclear plants, for this case in a nuclear power plant of ESBWR type. To achieve our objective it seeks to use specialized codes as RELAP/SCDAP to obtain simulations of passive autocatalytic recombiners behaviour and we can to estimate their operation inside the reactor contention, contemplating the possibility to use other codes like SCILAB and/or MATLAB for the simulation of a passive autocatalytic recombiner. (Author)

  12. Sustainability of common pool resources

    Science.gov (United States)

    Timilsina, Raja Rajendra; Kamijo, Yoshio

    2017-01-01

    Sustainability has become a key issue in managing natural resources together with growing concerns for capitalism, environmental and resource problems. We hypothesize that the ongoing modernization of competitive societies, which we refer to as “capitalism,” affects human nature for utilizing common pool resources, thus compromising sustainability. To test this hypothesis, we design and implement a set of dynamic common pool resource games and experiments in the following two types of Nepalese areas: (i) rural (non-capitalistic) and (ii) urban (capitalistic) areas. We find that a proportion of prosocial individuals in urban areas is lower than that in rural areas, and urban residents deplete resources more quickly than rural residents. The composition of proself and prosocial individuals in a group and the degree of capitalism are crucial in that an increase in prosocial members in a group and the rural dummy positively affect resource sustainability by 65% and 63%, respectively. Overall, this paper shows that when societies move toward more capitalistic environments, the sustainability of common pool resources tends to decrease with the changes in individual preferences, social norms, customs and views to others through human interactions. This result implies that individuals may be losing their coordination abilities for social dilemmas of resource sustainability in capitalistic societies. PMID:28212426

  13. Sustainability of common pool resources.

    Science.gov (United States)

    Timilsina, Raja Rajendra; Kotani, Koji; Kamijo, Yoshio

    2017-01-01

    Sustainability has become a key issue in managing natural resources together with growing concerns for capitalism, environmental and resource problems. We hypothesize that the ongoing modernization of competitive societies, which we refer to as "capitalism," affects human nature for utilizing common pool resources, thus compromising sustainability. To test this hypothesis, we design and implement a set of dynamic common pool resource games and experiments in the following two types of Nepalese areas: (i) rural (non-capitalistic) and (ii) urban (capitalistic) areas. We find that a proportion of prosocial individuals in urban areas is lower than that in rural areas, and urban residents deplete resources more quickly than rural residents. The composition of proself and prosocial individuals in a group and the degree of capitalism are crucial in that an increase in prosocial members in a group and the rural dummy positively affect resource sustainability by 65% and 63%, respectively. Overall, this paper shows that when societies move toward more capitalistic environments, the sustainability of common pool resources tends to decrease with the changes in individual preferences, social norms, customs and views to others through human interactions. This result implies that individuals may be losing their coordination abilities for social dilemmas of resource sustainability in capitalistic societies.

  14. The use of waveguide acoustic probes for void fraction measurement in the evaporator of BN-350-Type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Melnikov, V.I.; Nigmatulin, B.I.

    1995-09-01

    The present paper deals with some results of the experimental studies which have been carried out to investigate the steam generation dynamics in the Field tubes of sodium-water evaporators used in the BN-350 reactors. The void fraction measurements have been taken with the aid of waveguide acoustic transducers manufactured in accordance with a specially designed technology (waveguide acoustic transducers-WAT technology). Presented in this paper also the transducer design and calibration methods, as well as the diagram showing transducers arrengment in the evaporator. The transducers under test featured a waveguide of about 4 m in length and a 200-mm long sensitive element (probe). Besides, this paper specifies the void fraction data obtained through measurements in diverse points of the evaporator. The studies revealed that the period of observed fluctuations in the void fraction amounted to few seconds and was largely dependent on the level of water in the evaporator.

  15. Hydrogen Production in Fusion Reactors

    OpenAIRE

    Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M; Uenosono, C.

    1993-01-01

    As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated.

  16. Design of inventory pools in spare part support operation systems

    Science.gov (United States)

    Mo, Daniel Y.; Tseng, Mitchell M.; Cheung, Raymond K.

    2014-06-01

    The objective of a spare part support operation is to fulfill the part request order with different service contracts in the agreed response time. With this objective to achieve different service targets for multiple service contracts and the considerations of inventory investment, it is not only important to determine the inventory policy but also to design the structure of inventory pools and the order fulfilment strategies. In this research, we focused on two types of inventory pools: multiple inventory pool (MIP) and consolidated inventory pool (CIP). The idea of MIP is to maintain separated inventory pools based on the types of service contract, while CIP solely maintains a single inventory pool regardless of service contract. Our research aims to design the inventory pool analytically and propose reserve strategies to manage the order fulfilment risks in CIP. Mathematical models and simulation experiments would be applied for analysis and evaluation.

  17. IRL kutsub Ilvese esinema / Kalev Vilgats

    Index Scriptorium Estoniae

    Vilgats, Kalev

    2011-01-01

    Isamaa ja Res Publica Liidu eestseisus otsustas kutsuda president Toomas Hendrik Ilvese esinema 21. mail 2011 toimuvale volikogule, et otsustada tema kandidatuuri toetamise üle augustis toimuvatel presidendivalimistel

  18. Multifunctional reactors

    NARCIS (Netherlands)

    Westerterp, K.R.

    1992-01-01

    Multifunctional reactors are single pieces of equipment in which, besides the reaction, other functions are carried out simultaneously. The other functions can be a heat, mass or momentum transfer operation and even another reaction. Multifunctional reactors are not new, but they have received much

  19. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  20. Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

    Directory of Open Access Journals (Sweden)

    Kaiser Krista

    2016-01-01

    Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  1. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    Science.gov (United States)

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  2. Metal fires and their implications for advanced reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-10-01

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

  3. Vibro-acoustic Imaging at the Breazeale Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, James Arthur [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lee, James Edwin [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    The INL is developing Vibro-acoustic imaging technology to characterize microstructure in fuels and materials in spent fuel pools and within reactor vessels. A vibro-acoustic development laboratory has been established at the INL. The progress in developing the vibro-acoustic technology at the INL is the focus of this report. A successful technology demonstration was performed in a working TRIGA research reactor. Vibro-acoustic imaging was performed in the reactor pool of the Breazeale reactor in late September of 2015. A confocal transducer driven at a nominal 3 MHz was used to collect the 60 kHz differential beat frequency induced in a spent TRIGA fuel rod and empty gamma tube located in the main reactor water pool. Data was collected and analyzed with the INLDAS data acquisition software using a short time Fourier transform.

  4. The Future of Pooling.

    Science.gov (United States)

    Young, Peter C.; Fone, Martin

    1997-01-01

    Discusses seven propositions underlying the strategies that insurance pools can, will, and must pursue: (1) risk management versus risk financing; (2) elimination of windfall advantages; (3) the maintenance of market-dominant status; (4) cost leadership; (5) client focus; (6) innovation and diversification; and (7) leadership challenges. A sidebar…

  5. Income pooling within families

    DEFF Research Database (Denmark)

    Bonke, Jens; Uldall-Poulsen, Hans

    This paper analyses the phenomenon of income-pooling by applying the Danish household expenditure survey, merged with authoritative register information. Responses to additional questions on income sharing among 1696 couples also allows us to analyses whether the intra-household distribution of r...

  6. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  7. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  8. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Novog, D.R. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO{sub 2} in ThO{sub 2}) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO{sub 2} (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  9. Cold Pool and Surface Flux Interactions in Different Environments

    Science.gov (United States)

    Grant, L. D.; van den Heever, S. C.

    2015-12-01

    Cold pools play important roles in tropical and midlatitude deep convective initiation and organization through their influence on near-surface kinematic and thermodynamic fields. Because temperature, moisture, and winds are perturbed within cold pools, cold pools can also impact surface sensible and latent heat fluxes. In turn, surface fluxes both within the cold pool and in the environment can modify the characteristics of cold pools and their evolution, with subsequent implications for convective initiation and organization. The two-way interaction between cold pools and surface energy fluxes has not been well studied and is likely to vary according to the environment and surface type. The goal of this study is therefore to investigate the mechanisms by which surface fluxes and cold pools interact in environmental conditions ranging from tropical oceanic to dry continental. This goal will be accomplished using high-resolution (grid spacings as fine as 10 m), idealized, 2D simulations of isolated cold pools; such modeling experiments have proven useful for investigating cold pools and their dynamics in many previous studies. In the proposed experiments, the surface flux formulation, surface type, and environmental conditions will be systematically varied. The impact of surface fluxes on various cold pool characteristics and their evolution, including the buoyancy, maximum vertical velocity, and moisture distribution, will be analyzed and presented. Results suggest that the mechanisms by which surface fluxes and cold pools interact vary substantially with the environment. Additionally, the indirect effects of surface fluxes on turbulent entrainment rates into the cold pool are found to play an important role in cold pool evolution. These results suggest that surface fluxes can impact the timing and manner in which cold pools initiate convection, and that their effects may be important to incorporate into cold pool parameterizations for climate simulations.

  10. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: murakami@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Eguchi, Y., E-mail: eguchi@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Oyama, K., E-mail: kazuhiro_oyama@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan); Watanabe, O., E-mail: osamu4_watanabe@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan)

    2015-07-15

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  11. Helias reactor studies

    Energy Technology Data Exchange (ETDEWEB)

    Beidler, C.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Grieger, G. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Harmeyer, E. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Kisslinger, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Karulin, N. [Nuclear Fusion Institute, Moscow (Russian Federation); Maurer, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany); Nuehrenberg, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Rau, F. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Sapper, J. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Wobig, H. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1995-10-01

    The present status of Helias reactor studies is characterised by the identification and investigation of specific issues which result from the particular properties of this type of stellarator. On the technical side these are issues related to the coil system, while physics studies have concentrated on confinement, alpha-particle behaviour and ignition conditions. The usual assumptions have been made in those fields which are common to all toroidal fusion reactors: blanket and shield, refuelling and exhaust, safety and economic aspects. For blanket and shield sufficient space has been provided, a detailed concept will be developed in future. To date more emphasis has been placed on scoping and parameter studies as opposed to fixing a specific set of parameters and providing a detailed point study. One result of the Helias reactor studies is that physical dimensions are on the same order as those of tokamak reactors. However, it should be noticed that this comparison is difficult in view of the large spectrum of tokamak reactors ranging from a small reactor like Aries, to a large device such as SEAFP. The notion that the large aspect ratio of 10 or more in Helias configurations also leads to large reactors is misleading, since the large major radius of 22 m is compensated by the average plasma radius of 1.8 m and the average coil radius of 5 m. The plasma volume of 1400 m{sup 3} is about the same as the ITER reactor and the magnetic energy of the coil system is about the same or even slightly smaller than envisaged in ITER. (orig.)

  12. Anaerobic digestion of solid waste in RAS: Effect of reactor type on the biochemical acidogenic potential (BAP) and assessment of the biochemical methane potential (BMP) by a batch assay

    DEFF Research Database (Denmark)

    Suhr, Karin Isabel; Letelier-Gordo, Carlos Octavio; Lund, Ivar

    2015-01-01

    additional 14 and 20 days) in continuously stirred tank reactors. Generally, the VFA yield increased with time and no effect of the reactor type used was found within the time frame of the experiment. At 10 days HT or 10 days HRT the VFA yield reached 222.3 ± 30.5 and 203.4 ± 11.2 mg VFA g-1 TVS0 (total...... volatile solids at day 0) in batch and fed-batch reactor, respectively. For the fedbatch reactor, increasing HRT from 5 to 10 days gained no significant additional VFA yield. Prolonging the batch reactor experiment to 20 days increased VFA production further (273.9 ± 1.6 mg VFA g-1 TVS0, n=2). After 10...... for the design of an acidogenic continuously stirred reactor tank in a RAS single-sludge denitrification set-up. The biochemical methane potential of the sludge was estimated to 318 ± 29 g CH4 g-1 TVS0 by a batch assay and represented a higher utility of the solid waste when comparing the methane yield...

  13. Reactor vessel

    OpenAIRE

    Makkee, M.; Kapteijn, F.; Moulijn, J.A

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and comprising connections for supplying (9, 10, 11) and discharging (13, 14, 15) via the channels (3) gases and/or liquids entering into a reaction with each other and substances formed upon this reactio...

  14. Swimming pool use and birth defect risk.

    Science.gov (United States)

    Agopian, A J; Lupo, Philip J; Canfield, Mark A; Mitchell, Laura E

    2013-09-01

    Swimming during pregnancy is recommended. However, the use of swimming pools is also associated with infection by water-borne pathogens and exposure to water disinfection byproducts, which are 2 mechanisms that are suspected to increase risk for birth defects. Thus, we evaluated the relationship between maternal swimming pool use during early pregnancy and risk for select birth defects in offspring. Data were evaluated for nonsyndromic cases with 1 of 16 types of birth defects (n = 191-1829) and controls (n = 6826) from the National Birth Defects Prevention Study delivered during 2000-2006. Logistic regression analyses were conducted separately for each birth defect type. Separate analyses were conducted to assess any pool use (yes vs no) and frequent use (5 or more occasions in 1 month) during the month before pregnancy through the third month of pregnancy. There was no significant positive association between any or frequent pool use and any of the types of birth defects, even after adjustment for several potential confounders (maternal race/ethnicity, age at delivery, education, body mass index, folic acid use, nulliparity, smoking, annual household income, surveillance center, and season of conception). Frequent pool use was significantly negatively associated with spina bifida (adjusted odds ratio, 0.68; 95% confidence interval, 0.47-0.99). Among offspring of women 20 years old or older, pool use was associated with gastroschisis (adjusted odds ratio, 1.3; 95% confidence interval, 1.0-1.8), although not significantly so. We observed little evidence suggesting teratogenic effects of swimming pool use. Because swimming is a common and suggested form of exercise during pregnancy, these results are reassuring. Copyright © 2013 Mosby, Inc. All rights reserved.

  15. Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Wielgosz, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel is additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with

  16. Mixed convection and stratification phenomena in a heavy liquid metal pool

    Energy Technology Data Exchange (ETDEWEB)

    Tarantino, Mariano, E-mail: mariano.tarantino@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy); Martelli, Daniele; Barone, Gianluca [Dipartimento di Ingegneria Civile e Industriale, University of Pisa, Largo Lucio Lazzarino, 1-56100 Pisa Italy (Italy); Di Piazza, Ivan [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone (Italy); Forgione, Nicola [Dipartimento di Ingegneria Civile e Industriale, University of Pisa, Largo Lucio Lazzarino, 1-56100 Pisa Italy (Italy)

    2015-05-15

    Highlights: • Results related to experiments reproducing PLOHS + LOF accident in CIRCE pool facility. • Vertical thermal stratification in large HLM pool. • Transition from forced to natural circulation in HLM pool under DHR conditions. • Heat transfer coefficient measurement in HLM pin bundle. • Nusselt numbers calculations and comparison with correlations. - Abstract: This work deals with an analysis of the first experimental series of tests performed to investigate mixed convection and stratification phenomena in CIRCE HLM large pool. In particular, the tests concern the transition from nominal flow to natural circulation regime, typical of decay heat removal (DHR) regime. To this purpose the CIRCE pool facility has been updated to host a suitable test section in order to reproduce the thermal-hydraulic behaviour of a HLM pool-type reactor. The test section basically consists of an electrical bundle (FPS) made up of 37 pins arranged in a hexagonal wrapped lattice with a pitch diameter ratio of 1.8. Along the FPS active length, three sections were instrumented to monitor the heat transfer coefficient along the bundle as well as the cladding temperatures at different ranks of the sub-channels. This paper reports the experimental data as well as a preliminary analysis and discussion of the results, focusing on the most relevant tests of the campaign, namely Test I (48 h) and Test II (97 h). Temperatures along three sections of the FPS and at inlet and outlet sections of the main components were reported and the Nusselt number in the FPS sub-channels was investigated together with the void fraction in the riser. Concerning the investigation of in-pool thermal stratification phenomena, the temperatures in the whole LBE pool were monitored at different elevations and radial locations. The analysis of experimental data obtained from Tests I and II underline the occurrence of thermal stratification phenomena in the region placed between the outlet sections of

  17. Alpha-glucosidase inhibitor, acarbose, improves glycamic control and reduces body weight in type 2 diabetes: Findings on indian patients from the pooled data analysis

    Directory of Open Access Journals (Sweden)

    Sanjay Kalra

    2013-01-01

    Full Text Available Alpha-glucosidase inhibitors are widely used especially in Asian countries as a treatment option for type 2 diabetes patients with high postprandial glycemia (PPG. The higher carbohydrate in the Indian diets lead to greater prandial glycemic excursion, increased glucosidase, and incretin activity in the gut and may need special therapeutic strategies to tackle these glucose peaks. This is the subgroup analysis of Indian subjects who participated in the GlucoVIP study that investigated the effectiveness and tolerability of acarbose as add-on or monotherapy in a range of patients with type 2 diabetes mellitus. A total of 1996 Indian patients were included in the effectiveness analysis. After 12.5 weeks (mean, the mean change in 2-hour PPG from baseline was −74.4 mg/dl, mean HbA1c decreased by -1.0%, and mean fasting blood glucose decreased by -37.9 mg/dl. The efficacy of acarbose was rated "very good" or "good" in 91.1% of patients, and tolerability as "very good" or "good" in 88.0% of patients. The results of this observational study suggest that acarbose was effective and well tolerated in the Indian patients with T2DM.

  18. Chemical Reactors.

    Science.gov (United States)

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  19. Reactor Neutrinos

    Directory of Open Access Journals (Sweden)

    Soo-Bong Kim

    2013-01-01

    Full Text Available We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very recently the most precise determination of the neutrino mixing angle θ13. This paper provides an overview of the upcoming experiments and of the projects under development, including the determination of the neutrino mass hierarchy and the possible use of neutrinos for society, for nonproliferation of nuclear materials, and geophysics.

  20. Reactor Engineering

    Science.gov (United States)

    Lema, Juan M.; López, Carmen; Eibes, Gemma; Taboada-Puig, Roberto; Moreira, M. Teresa; Feijoo, Gumersindo

    In this chapter, the engineering aspects of processes catalyzed by peroxidases will be presented. In particular, a discussion of the existing technologies that utilize peroxidases for different purposes, such as the removal of recalcitrant compounds or the synthesis of polymers, is analyzed. In the first section, the essential variables controlling the process will be investigated, not only those that are common in any enzymatic system but also those specific to peroxidative reactions. Next, different reactor configurations and operational modes will be proposed, emphasizing their suitability and unsuitability for different systems. Finally, two specific reactors will be described in detail: enzymatic membrane reactors and biphasic reactors. These configurations are especially valuable for the treatment of xenobiotics with high and poor water solubility, respectively.

  1. Reactor Neutrinos

    OpenAIRE

    Lasserre, T.; Sobel, H.W.

    2005-01-01

    We review the status and the results of reactor neutrino experiments, that toe the cutting edge of neutrino research. Short baseline experiments have provided the measurement of the reactor neutrino spectrum, and are still searching for important phenomena such as the neutrino magnetic moment. They could open the door to the measurement of coherent neutrino scattering in a near future. Middle and long baseline oscillation experiments at Chooz and KamLAND have played a relevant role in neutrin...

  2. The reactor Cabri; La pile cabri

    Energy Technology Data Exchange (ETDEWEB)

    Ailloud, J.; Millot, J.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m{sup 3}/h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under

  3. CERN Electronics Pool presentations

    CERN Multimedia

    2011-01-01

    The CERN Electronics Pool has organised a series of presentations in collaboration with oscilloscope manufacturers. The last one will take place according to the schedule below.   Time will be available at the end of the presentation to discuss your personal needs. The Agilent presentation had to be postponed and will be organised later. -     Lecroy: Thursday, 24 November 2011, in 530-R-030, 14:00 to 16:30.

  4. Heavy Water Reactor; Reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Yu, St.; HOpwood, J.; Meneley, D. [Energie Atomique du Canada (Canada)

    2000-04-01

    This document deals with the Heavy Water Reactor (HWR) technology and especially the Candu (Canada Deuterium Uranium) reactor. This reactors type offers many advantages that promote them for the future. General concepts, a description of the Candu nuclear power plants, the safety systems, the fuel cycle and economical and environmental aspects are included. (A.L.B.)

  5. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature.

  6. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    Science.gov (United States)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  7. Claim criteria of significant events implying the safety of PWR type reactors; Criteres de declaration des evenements significatifs impliquant la surete pour les reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-10-15

    There are ten criteria for the declaration of the significant events implying the safety for PWR type reactors. First criterion: Automatic stop of the reactor: manual or automatic, inconvenient starting or not, the function of automatic stop of the reactor, whatever is the state of the reactor, with the exception of the deliberate starting resulting from planned actions. Second criterion: Starting of one of the systems of protection, manual or automatic, inconvenient starting or not, of one of the systems of protection, with the exception of the deliberate starting resulting from planned actions. Third criterion: Disregard of the technical specifications of exploitation (S.T.E ), or an event which would have been able to lead to a disregard of the S.T.E., if the same event had occurred, the installation having been in a different state, any disregard of one or several permanent conditions defined in S.T.E., any disregard of the conditions of a dispensation in S.T.E., any overtaking of periods when it is not prescribed by state of fold, any unavailability provoked outside the conditions planned by the main rules of exploitation, not identified beforehand or identified but untreated according to the prescriptions of the S.T.E. fourth criterion: Internal or external aggression, happening of a natural external phenomenon or in relation with a human activity, or happening of an internal flooding, a fire or another phenomenon susceptible to affect the availability of the equipment important for the safety. Fifth criterion: Act or attempt of act of hostility susceptible to affect the safety of the installation. Sixth criterion: Passage in state of fold in application of the technical specifications of exploitation or the accidental procedures of driving following an unforeseen behavior of the installation. Seventh criterion: Event having cause or being able to cause multiple failures, unavailability of equipment due to the same failure either affecting all the ways of a

  8. Dominant factors in controlling marine gas pools in South China

    Science.gov (United States)

    Xu, S.; Watney, W.L.

    2007-01-01

    In marine strata from Sinian to Middle Triassic in South China, there develop four sets of regional and six sets of local source rocks, and ten sets of reservoir rocks. The occurrence of four main formation periods in association with five main reconstruction periods, results in a secondary origin for the most marine gas pools in South China. To improve the understanding of marine gas pools in South China with severely deformed geological background, the dominant control factors are discussed in this paper. The fluid sources, including the gas cracked from crude oil, the gas dissolved in water, the gas of inorganic origin, hydrocarbons generated during the second phase, and the mixed pool fluid source, were the most significant control factors of the types and the development stage of pools. The period of the pool formation and the reconstruction controlled the pool evolution and the distribution on a regional scale. Owing to the multiple periods of the pool formation and the reconstruction, the distribution of marine gas pools was complex both in space and in time, and the gas in the pools is heterogeneous. Pool elements, such as preservation conditions, traps and migration paths, and reservoir rocks and facies, also served as important control factors to marine gas pools in South China. Especially, the preservation conditions played a key role in maintaining marine oil and gas accumulations on a regional or local scale. According to several dominant control factors of a pool, the pool-controlling model can be constructed. As an example, the pool-controlling model of Sinian gas pool in Weiyuan gas field in Sichuan basin was summed up. ?? Higher Education Press and Springer-Verlag 2007.

  9. Dominant factors in controlling marine gas pools in South China

    Institute of Scientific and Technical Information of China (English)

    XU Sihuang; W.Lynn Watney

    2007-01-01

    In marine strata from Sinian to Middle Triassic in South China,there develop four sets of regional and six sets of local source rocks,and ten sets of reservoir rocks.The occurrence of four main formation periods in association with five main reconstruction periods,results in a secondary origin for the most marine gas pools in South China.To improve the understanding of marine gas pools in South China with severely deformed geological background,the dominant control factors are discussed in this paper.The fluid sources,including the gas cracked from crude oil,the gas dissolved in water,the gas of inorganic origin,hydrocarbons generated during the second phase,and the mixed pool fluid source,were the most significant control factors of the types and the development stage of pools.The period of the pool formation and the reconstruction controlled the pool evolution and the distribution on a regional scale.Owing to the multiple periods of the pool formation and the reconstruction,the distribution of marine gas pools was complex both in space and in time,and the gas in the pools is heterogeneous.Pool elements,such as preservation conditions,traps and migration paths,and reservoir rocks and facies,also served as important control factors to marine gas pools in South China.Especially,the preservation conditions played a key role in maintaining marine oil and gas accumulations on a regional or local scale.According to several dominant control factors of a pool,the pool-controlling model can be constructed.As an example,the pool-controlling model of Sinian gas pool in Weiyuan gas field in Sichuan basin was summed up.

  10. Mesophilic anaerobic digestion of several types of spent livestock bedding in a batch leach-bed reactor: substrate characterization and process performance.

    Science.gov (United States)

    Riggio, S; Torrijos, M; Debord, R; Esposito, G; van Hullebusch, E D; Steyer, J P; Escudié, R

    2017-01-01

    Spent animal bedding is a valuable resource for green energy production in rural areas. The properties of six types of spent bedding collected from deep-litter stables, housing either sheeps, goats, horses or cows, were compared and their anaerobic digestion in a batch Leach-Bed Reactor (LBR) was assessed. Spent horse bedding, when compared to all the other types, appeared to differ the most due to a greater amount of straw added to the litter and a more frequent litter change. Total solids content appeared to vary significantly from one bedding type to another, with consequent impact on the methane produced from the raw substrate. However, all the types of spent bedding had similar VS/TS (82.3-88.9)%, a C/N well-suited to anaerobic digestion (20-28, except that of the horse, 42) and their BMPs were in a narrow range (192-239NmLCH4/gVS). The anaerobic digestion in each LBR was stable and the pH always remained higher than 6.6 regardless of the type of bedding. In contrast to all the other substrates, spent goat bedding showed a stronger acidification resulting in a methane production lag phase. Finally, spent bedding of different origins reached, on average, (89±11)% of their BMP after 60days of operation. This means that this waste is well-suited for treatment in LBRs and that this is a promising process to recover energy from dry agricultural waste. Copyright © 2016 Elsevier Ltd. All rights reserved.

  11. ROLE OF PASSIVE SAFETY FEATURES IN PREVENTION AND MITIGATION OF SEVERE PLANT CONDITIONS IN INDIAN ADVANCED HEAVY WATER REACTOR

    Directory of Open Access Journals (Sweden)

    VIKAS JAIN

    2013-10-01

    Full Text Available Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor ‘AHWR’ is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI, Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

  12. Efficacy and safety of canagliflozin compared with placebo in older patients with type 2 diabetes mellitus: a pooled analysis of clinical studies

    Science.gov (United States)

    2014-01-01

    Background Canagliflozin is a sodium glucose co-transporter 2 inhibitor developed for the treatment of patients with type 2 diabetes mellitus (T2DM). The efficacy and safety of canagliflozin were evaluated in patients with T2DM canagliflozin 100 and 300 mg were analysed by age: Canagliflozin 100 and 300 mg reduced HbA1c and fasting plasma glucose relative to placebo in patients canagliflozin doses reduced body weight and systolic BP relative to placebo in patients canagliflozin 100 mg than other groups in patients ≥65 years of age. As in patients canagliflozin relative to placebo in those ≥65 years of age. Incidences of urinary tract infections (UTIs), renal-related AEs, AEs related to volume depletion, and documented hypoglycaemia episodes were similar across all treatment groups in patients ≥65 years of age; no notable trends were observed with canagliflozin 100 and 300 mg relative to placebo in these AEs among patients canagliflozin were similar in both age subsets. Conclusions Canagliflozin improved glycaemic control, body weight, and systolic BP, and was generally well tolerated in older patients with T2DM. Trial registration ClinicalTrials.gov, NCT01081834; NCT01106677; NCT01106625; NCT01106690. PMID:24742013

  13. Efficacy and safety of empagliflozin in Japanese patients with type 2 diabetes mellitus: A sub-analysis by body mass index and age of pooled data from three clinical trials.

    Science.gov (United States)

    Shiba, Teruo; Ishii, So; Okamura, Tomoo; Mitsuyoshi, Rika; Pfarr, Egon; Koiwai, Kazuki

    2017-09-01

    To investigate the efficacy and safety of empagliflozin in subgroups based on body mass index (BMI) and age, using a pooled data set from Japanese patients with type 2 diabetes mellitus (T2DM). Pooled data from 1403 patients treated with empagliflozin at 10mg/day or 25mg/day in three clinical studies (≥52week treatment) were stratified by baseline BMI (<22, 22 to <25 and ≥25kg/m(2)) and baseline age (<50, 50 to <65 and ≥65years). Empagliflozin at 10mg/day and 25mg/day reduced mean glycated hemoglobin (HbA1c) (-0.77 to -0.87% and -0.76 to -0.97%, respectively), mean fasting plasma glucose (FPG) (-20.79 to -27.06mg/dL and -26.08 to -29.60mg/dL) and mean body weight (-3.4 to -4.7% and -3.7 to -4.7%) in all subgroups of baseline BMI and age, regardless of age and degree of obesity. Adverse events were observed in approximately 70-80% patients in BMI and age subgroups of both empagliflozin groups. No hypoglycemia requiring assistance was observed. Neither UTI nor genital infection rates differed markedly among the BMI and age subgroups. Volume depletion was increased in patients ≥65years of age as compared to younger patients. Empagliflozin was well tolerated and improved HbA1c, FPG and body weight in all BMI and age subgroups of Japanese patients with T2DM, regardless of age and degree of obesity. Empagliflozin is considered to be effective and well tolerated for treating a wide range of Japanese patients with T2DM. Study 1 (NCT01193218), Study 2 (NCT01289990) and Study 3 (NCT01368081). Copyright © 2017 The Authors. Published by Elsevier B.V. All rights reserved.

  14. Blood pressure and fasting lipid changes after 24 weeks’ treatment with vildagliptin: a pooled analysis in >2,000 previously drug-naïve patients with type 2 diabetes mellitus

    Science.gov (United States)

    Evans, Marc; Schweizer, Anja; Foley, James E

    2016-01-01

    Introduction We have previously shown modest weight loss with vildagliptin treatment. Since body weight balance is associated with changes in blood pressure (BP) and fasting lipids, we have assessed these parameters following vildagliptin treatment. Methods Data were pooled from all double-blind, randomized, controlled, vildagliptin mono-therapy trials on previously drug-naïve patients with type 2 diabetes mellitus who received vildagliptin 50 mg once daily (qd) or twice daily (bid; n=2,108) and wherein BP and fasting lipid data were obtained. Results Data from patients receiving vildagliptin 50 mg qd or bid showed reductions from baseline to week 24 in systolic BP (from 132.5±0.32 to 129.8±0.34 mmHg; P<0.0001), diastolic BP (from 81.2±0.18 to 79.6±0.19 mmHg; P<0.0001), fasting triglycerides (from 2.00±0.02 to 1.80±0.02 mmol/L; P<0.0001), very low density lipoprotein cholesterol (from 0.90±0.01 to 0.83±0.01 mmol/L; P<0.0001), and low density lipoprotein cholesterol (from 3.17±0.02 to 3.04±0.02 mmol/L; P<0.0001), whereas high density lipoprotein cholesterol increased (from 1.19±0.01 to 1.22±0.01 mmol/L; P<0.001). Weight decreased by 0.48±0.08 kg (P<0.001). Conclusion This large pooled analysis demonstrated that vildagliptin shows a significant reduction in BP and a favorable fasting lipid profile that are associated with modest weight loss. PMID:27574437

  15. Steam-air mixture condensation in a subcooled water pool

    Science.gov (United States)

    Norman, Timothy Linhurst

    2007-12-01

    In any conceptual reactor design under postulated accidental conditions, one parameter that is considered as being highly ranked in determining the thermal-hydraulic conditions of the reactor safety components is the system pressure. To obtain a satisfactory prediction of steam partial pressure, within reasonable uncertainty in the gas space of a confined SP (suppression pool) bounded to the steam source of the break flow, one must establish a means by which local phenomena associated with steam direct contact condensation in the subcooled water pool can be fully addressed to predict the global component thermal response. For this purpose a scaled down, reduced pressure, suppression pool was designed and built to study condensation and mixing phenomena. The scaled test facility represented an idealized trapezoidal cross section, 1/10 sector of the SP with scaled height ratio of 1/4.5 and volume ratio of 1/400. The design and test conditions were based on a hierarchical scaling principle that preserves the transfer of mass, momentum, energy and condensation phenomena. Distributed thermocouples within the pool provided a means to quantify the pool thermal response. The test loop was not only instrumented with thermocouples for monitoring pool stratification but also with high speed photography for flow visualization from which to build a comprehensive database to identify the regions of the pool that were thermally stratified or mixed. Data were obtained for different pool initial subcooling and steam/air mixture flow rates. Dimensionless boundary maps were plotted from several experimental runs of pure steam injection to determine conditions when the pool transits from being homogeneously mixed to being thermally stratified. Steam-air mixture injection cases for single horizontal venting indicated that above a pool temperature of 40°C with airmass flow rates below 0.1 g/s the pool can attain thermal stratification. Models of a single phase liquid

  16. Swimming Pools and Molluscum Contagiosum

    Science.gov (United States)

    ... Travelers’ Health: Smallpox & Other Orthopoxvirus-Associated Infections Poxvirus Swimming Pools Recommend on Facebook Tweet Share Compartir The ... often ask if molluscum virus can spread in swimming pools. There is also concern that it can ...

  17. Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

    Energy Technology Data Exchange (ETDEWEB)

    Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

  18. Preliminary Calculation on a Spent Fuel Pool Accident using GOTHIC

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaehwan; Choi, Yu Jung; Hong, Tae Hyub; Kim, Hyeong-Taek [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The probability of an accident happening at the spent fuel pool was believed to be quite low until the 2011 Fukushima accident occurred. Notably, large amount of spent fuel are normally stored in the spent fuel pool for a long time compared to the amount of fuel in the reactor core and the total heat released from the spent fuel is high enough to boil the water of the spent fuel pool when the cooling system does not operate. In addition, the enrichment and the burnup of the fuel have both increased in the past decade and heat generation from the spent fuel thereby has also increased. The failure of the cooling system at the spent fuel pool (hereafter, a loss-of-cooling accident) is one of the principal hypothetical causes of an accident that could occur at the spent fuel pool. In this paper, the preliminary calculation of a loss-of-cooling accident was performed. In this paper, the preliminary calculation of a loss-of cooling accident was performed with GOTHIC. The calculation results show boiling away of water in the spent fuel pool due to the loss-of-cooling accident and similar thermal performance of the spent fuel pool with previous research results.

  19. Relationships between obesity, glycemic control, and cardiovascular risk factors: a pooled analysis of cross-sectional data from Spanish patients with type 2 diabetes in the preinsulin stage.

    Science.gov (United States)

    Vázquez, Luis A; Rodríguez, Ángel; Salvador, Javier; Ascaso, Juan F; Petto, Helmut; Reviriego, Jesús

    2014-11-01

    Obesity is associated with the onset of type 2 diabetes mellitus (T2D), but reports conflict regarding the association between obesity and macrovascular complications. In this study, we investigated associations between cardiovascular risk factors and body mass index (BMI) and glycemic control in non-insulin-treated patients with T2D. Authors gathered cross-sectional data from five observational studies performed in Spain. Generalized logit models were used to analyze the relationship between cardiovascular risk factors (independent variables) and 5 BMI strata (6.5-7%, >7-8%, >8-9%, >9%) (dependent outcomes). In total, data from 6442 patients were analyzed. Patients generally had mean values of investigated cardiovascular risk factors outside recommended thresholds. Younger patients had higher BMI, triglyceride levels and HbA1c than their older counterparts. Diastolic blood pressure, systolic blood pressure and triglyceride levels were directly correlated with BMI strata, whereas an inverse correlation was observed between BMI strata and high-density lipoprotein cholesterol (HDL-C) levels, patient age, and duration of T2D. Increased duration of T2D and total cholesterol levels, and decreased HDL-C levels were associated with a higher HbA1c category. BMI and HbA1c levels were not associated with each other. As insulin-naïve patients with T2D became more obese, cardiovascular risk factors became more pronounced. Higher BMI was associated with younger age and shorter duration of T2D, consistent with the notion that obesity at an early age may be key to the current T2D epidemic. Glycemic control was independent of BMI but associated with abnormal lipid levels. Further efforts should be done to improve modifiable cardiovascular risk factors.

  20. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    Currently, the lifetime of control rods in JOYO is limited by Absorber-Cladding Mechanical Interaction (ACMI) due to swelling of B{sub 4}C(boron carbide) pellets accelerated by relocation of pellet fragments. A sodium bonded type control rod was developed which improves the thermal conductivity by means of charging sodium into the gap between B{sub 4}C and cladding and by utilizing a shroud which wraps the pellet fragments in a thin tube. This new design will be able to enlarge the gap between B{sub 4}C and cladding, without heating B{sub 4}C or fragment relocation, thus extending the life of the control rod. The sodium bonded type will be fabricated as the ninth reload control rods in JOYO. (1) The specification of a sodium bonded type control rod was determined with the wide gap between B{sub 4}C and cladding. In the design simulation, main component temperature were below the maximum limit. And the local heating by helium bubble generated from B{sub 4}C in the sodium gap, was not a serious problem in the analysis which was considered. (2) A structural design for the sodium entrance into the pin was determined. A formula was developed which the limit for sodium charging given physical dimension of the structure and sodium property. Result from sodium out-pile experiments validated the theoretical formula. (3) The analysis of ACMI indicated a lifetime extension of the sodium bonded type by 4.6% in comparison with lifetime of the helium bonded type of 1.6%. This is due to the boron10 burn-up rate being three times higher in the sodium bonded type than in the helium bonded type. To achieve a target burn-up 10% in the future, it will be necessary to modify design based on irradiation data which will be obtained by practical use of the sodium bonded control rods in JOYO. (4) The effects due to Absorber-Cladding Chemical Interaction (ACCI) were reduced by controlling the cladding temperature and chromium coating to the cladding's inner surface. It was confirmed

  1. Efficacy and safety of linagliptin in type 2 diabetes subjects at high risk for renal and cardiovascular disease: a pooled analysis of six phase III clinical trials.

    Science.gov (United States)

    von Eynatten, Maximilian; Gong, Yan; Emser, Angela; Woerle, Hans-Juergen

    2013-04-09

    In patients with type 2 diabetes mellitus (T2DM), hypertension and microalbuminuria are predictive markers for increased renal and cardiovascular risk. This post hoc analysis of data from a global development program aimed to evaluate the efficacy and safety of linagliptin in a population with joint prevalence of these two vascular risk factors. Data for patients with baseline microalbuminuria (urine albumin-to-creatinine ratio 30-300 mg/g) and hypertension (systolic blood pressure ≥ 140 mm Hg and/or diastolic blood pressure ≥ 90 mm Hg and/or a history of hypertension; and/or an antihypertensive treatment at baseline) who participated in any of six randomized, placebo-controlled, phase III trials were analyzed. Participants received linagliptin 5 mg daily (alone or in combination with other oral antidiabetic drugs) or placebo for 18 to 24 weeks. Of 3,119 patients, 512 had both microalbuminuria and hypertension (linagliptin, 366; placebo, 146). Baseline mean (SD) HbA1c was 8.3 (0.9)% and 8.4 (0.9)%; median (range) urine albumin-to-creatinine ratio was 60 (30-292) mg/g and 64 (30-298) mg/g; mean (SD) systolic blood pressure was 138 (15) mm Hg and 135 (16) mm Hg; and mean (SD) diastolic blood pressure was 81 (10) mm Hg and 81 (10) mm Hg, for linagliptin and placebo, respectively. Placebo-corrected mean change in HbA1c from baseline to week 18 and week 24 was -0.57% (95% CI: -0.75, -0.39; P blood pressure, cholesterol and triglyceride levels were similar between linagliptin and placebo. In T2DM patients with the two common vascular risk factors of hypertension and microalbuminuria, linagliptin achieved significant improvements in glycemic control. In this vulnerable patient population at high risk for micro- and macrovascular complications, linagliptin was well tolerated.

  2. Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

    Science.gov (United States)

    Ioltukhovskiy, A. G.; Leonteva-Smirnova, M. V.; Solonin, M. I.; Chernov, V. M.; Golovanov, V. N.; Shamardin, V. K.; Bulanova, T. M.; Povstyanko, A. V.; Fedoseev, A. E.

    2002-12-01

    Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a δ-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 °C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 °C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

  3. Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ioltukhovskiy, A.G. E-mail: iral@bochvar.ru; Leonteva-Smirnova, M.V.; Solonin, M.I.; Chernov, V.M.; Golovanov, V.N.; Shamardin, V.K.; Bulanova, T.M.; Povstyanko, A.V.; Fedoseev, A.E

    2002-12-01

    Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a {delta}-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 deg. C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 deg. C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

  4. Mathematical model analysis on the enhancement of aeration efficiency using ladder-type flat membrane module forms in the Submerged Membrane Bio-reactor(SMBR)

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    The cross-flow shearing action produced from the inferior aeration in the Submerged Membrane Bio-reactor(SMBR) is an effective way to further improve anti-fouling effects of membrane modules.Based on the widely-applied vertical structure of flat membrane modules,improvements are made that ladder-type flat membrane structure is designed with a certain inclined angle θ so that the cross-flow velocity of bubble near the membrane surface can be held,and the intensity and times of elastic colli-sion between bubbles and membrane surface can be increased.This can improve scouring action of membrane surface on aeration and reduce energy consumption of strong aeration in SMBR.By de-ducing and improving the mathematics model of collision between bubble and vertical flat put forward by Vries,the relatively suitable incline angle θ under certain aeration place and in certain size rang of bubble can be obtained with the computer iterative calculation technology.Finally,for many groups of ladder-type flat membrane in parallel placement in the practical application of SMBR,some sugges-tions are offered:the interval distance of membrane modules is 8―15 mm,and aeration should be op-erated at 5―7 mm among membrane modules,and the optimal design angle of trapeziform membrane is 1.7°―2.5°.

  5. Mathematical model analysis on the enhancement of aeration efficiency using ladder-type flat membrane module forms in the Submerged Membrane Bio-reactor (SMBR)

    Institute of Scientific and Technical Information of China (English)

    LI Bo; YE MaoSheng; YANG FengLin; MA Hui

    2009-01-01

    The cross-flow shearing action produced from the inferior aeration in the Submerged Membrane Bio-reactor (SMBR) Is an effective way to further improve anti-fouling effects of membrane modules.Based on the widely-applied vertical structure of flat membrane modules, improvements are made that ladder-type flat membrane structure is designed with a certain inclined angle θ so that the cross-flow velocity of bubble near the membrane surface can be held, and the intensity and times of elastic colli-sion between bubbles and membrane surface can be increased. This can improve scouring action ofmembrane surface on aeration and reduce energy consumption of strong aeration in SMBR. By de-ducing and improving the mathematics model of collision between bubble and vertical flat put forward by Vries, the relatively suitable Incline angle θ under certain aeration place and in certain size rang ofbubble can be obtained with the computer iterative calculation technology. Finally, for many groups of ladder-type flat membrane in parallel placement in the practical application of SMBR, some sugges-tions are offered: the interval distance of membrane modules is 8--15 mm, and aeration should be op-erated at 5--7 mm among membrane modules, and the optimal design angle of trapeziform membrane is 1.7°--2.5°.

  6. 24 CFR 320.9 - Pool administration.

    Science.gov (United States)

    2010-04-01

    ...) GOVERNMENT NATIONAL MORTGAGE ASSOCIATION, DEPARTMENT OF HOUSING AND URBAN DEVELOPMENT GUARANTY OF MORTGAGE-BACKED SECURITIES Pass-Through Type Securities § 320.9 Pool administration. The Association will only guarantee securities if the issuer executes a guaranty agreement or contractual agreement in the form...

  7. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  8. Use of gadolinium burnable absorbers in VVER Type Reactors. Validation of WIMS-D/4 code; Empleo del gadolinio como absorbente quemable en los reactores nucleares VVER. Validacion del codigo WIMS-D/4

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez Cardona, Caridad M.; Guerra Valdes, Ramiro; Lopez Aldama, Daniel [Centro de Tecnologia Nuclear, La Habana (Cuba)

    1996-07-01

    Burnable absorbers are not used in current operating WWERs, but in order to optimize the fuel cycle and enhance operational safety, one should also introduce gadolinium or a similar burnable absorber in these reactors. For this purpose adequate tools for properly calculating local effects in hexagonal geometries should be developed and validated. The present gives main results in validating the WIMS-D/4 lattice code for Gd burnable absorber bearing WWER lattices. To validate the code experimental and calculational benchmarks proposed in a IAEA Coordinated Research Program were solved. A code system for the optimization of the Gd axial distribution in a WWER reactor was developed and it also presented here. (author)

  9. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  10. The Productive Ligurian Pool

    CERN Document Server

    Casella, E; Couvelard, X; Caldeira, R M A

    2011-01-01

    In contrast with the behavior of the eddies in the open-ocean, the sub-mesoscale eddies generated in the constricted Ligurian Basin (NW Mediterranean), are unproductive but their combined effect, arranged in a rim-like fashion, contributes to the containment of a Productive Ligurian Pool (PLP). Data de- rived from MODIS satellite sensor showed persistent higher chlorophyll con- centrations in the centre of the basin, concurrent with high EKE values in its surroundings, derived from AVISO altimetry merged products. This sug- gested that this 'productive pool' is maintained by the intense (sub)mesoscale eddy activity in the rim. Numerical realistic experiments, using a Regional Ocean Model System, forced by MERCATOR and by a high-resolution COSMO- l7 atmospheric model, also showed that most of the sub-mesoscale eddies, during 2009 and 2010, are concentrated in the rim surrounding the basin, contributing to the formation of a basin-scale cyclonic gyre. We hypothesized that the interaction between eddies in the r...

  11. Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

    Energy Technology Data Exchange (ETDEWEB)

    Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

    2015-12-01

    BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

  12. Progress of China Experimental Fast Reactor in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    1 Background Fast reactor is the reactor which realized the chain fission with fast neutron.As an optional type of generation Ⅳ reactor,fast reactor has three characters:1) It can change 238U to 239Pu and raise the uranium resource utilization

  13. Antineutrino Monitoring of Thorium Reactors

    CERN Document Server

    Akindele, Oluwatomi A; Norman, Eric B

    2015-01-01

    Various groups have demonstrated that antineutrino monitoring can be successful in assessing the plutonium content in water-cooled nuclear reactors for nonproliferation applications. New reactor designs and concepts incorporate nontraditional fuels types and chemistry. Understanding how these properties affect the antineutrino emission from a reactor can extend the applicability of antineutrino monitoring.Thorium molten salt reactors (MSR) breed U-233, that if diverted constitute an IAEA direct use material. The antineutrino spectrum from the fission of U-233 has been determined, the feasibility of detecting the diversion of a significant quantity, 8 kg of U-233, within the IAEA timeliness goal of 30 days has been evaluated. The antineutrino emission from a thorium reactor operating under normal conditions is compared to a diversion scenario at a 25 meter standoff by evaluating the daily antineutrino count rate and the energy spectrum of the detected antineutrinos. It was found that the diversion of a signifi...

  14. Serotonin storage pools in basophil leukemia and mast cells: characterization of two types of serotonin binding protein and radioautographic analysis of the intracellular distribution of (/sup 3/H)serotonin

    Energy Technology Data Exchange (ETDEWEB)

    Tamir, H. (New York Psychiatric Inst., New York); Theoharides, T.C.; Gershon, M.D.; Askenase, P.W.

    1982-06-01

    The binding of serotonin to protein(s) derived from rat basophil leukemia (RBL) cells and mast cells was studied. Two types of serotonin binding protein in RBL cells was found. These proteins differed from one another in molecular weight and eluted in separate peaks from sephadex G-200 columns. Peak I protein (KD = 1.9 x 10/sup -6/ M) was a glycoprotein that bound to concanavalin A (Con A); Peak II protein (KD/sub 1/ = 4.5 x 10/sup -/8 M; KD/sub 2/ = 3.9 x 10/sup -6/ M) did not bind to Con A. Moreover, binding of (/sup 3/H)serotonin to protein of Peak I was sensitive to inhibition by reserpine, while binding of (/sup 3/H)serotonin to protein of Peak II resisted inhibition by that drug. Other differences between the two types of binding protein were found, the most significant of which was the far more vigorous conditions of homogenization required to extract Peak I than Peak II protein. Electron microscope radioautographic analysis of the intracellular distribution of (/sup 3/H) serotonin taken up in vitro by RBL cells or in vivo by murine mast cells indicated that essentially all of the labeled amine was located in cytoplasmic granules.No evidence for a pool in the cytosol was found and all granules were capable of becoming labeled. The presence of two types of intracellular serotonin binding proteins in these cells may indicate that there are two intracellular storage compartments for the amine. Both may be intragranular, but Peak I protein may be associated with the granular membrane while Peak II protein may be more free within the granular core. Different storage proteins may help to explain the differential release of amines from mast cell granules.

  15. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2010-12-15

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  16. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor; Estimacion de la fluencia de neutrones rapidos en probetas de acero tipo Laguna Verde en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO{sub 3}) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10{sup 18} n/cm{sup 2}, which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  17. Decay profiles of {beta} and {gamma} for a radionuclide inventory in equilibrium cycle of a BWR type reactor; Perfiles de decaimiento de radiacion {beta} y {gamma} para un inventario de radionuclidos en ciclo de equilibrio de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Salaices, M.; Sandoval, S.; Ovando, R. [Instituto de Investigaciones Electricas. Gerencia de Energia Nuclear, Av. Reforma 113 Col. Palmira. 62490 Cuernavaca, Morelos (Mexico)]. e-mail: sal@iie.org.mx

    2007-07-01

    Presently work the {beta} and {gamma} radiation decay profiles for a radionuclides inventory in equilibrium cycle of a BWR type reactor is presented. The profiles are presented in terms of decay in the activity of the total inventory as well as of the chemical groups that conform the inventory. In the obtaining of the radionuclides inventory in equilibrium cycle the ORIGEN2 code, version 1 was used, which simulates fuel burnup cycles and it calculates the evolution of the isotopic composition as a result of the burnt one, irradiation and decay of the nuclear fuel. It can be observed starting from the results that the decrease in the activity for the initial inventory and the different chemical groups that conform it is approximately proportional to the base 10 logarithm of the time for the first 24 hours of having concluded the burnt one. It can also be observed that the chemical groups that contribute in more proportion to the total activity of the inventory are the lanthanides-actinides and the transition metals, with 39% and 28%, respectively. The groups of alkaline earth metals, halogens, metalloids, noble gases and alkaline metals, contribute with percentages that go from the 8 to 5%. The groups that less they contribute to the total activity of the inventory they are the non metals and semi-metals with smaller proportions that 1%. The chemical groups that more contribute to the energy of {beta} and {gamma} radiation its are the transition metals and the lanthanides-actinides with a change in the order of importance at the end of the 24 hours period. The case of the halogens is of relevance for the case of the {gamma} radiation energy due that occupying the very near third site to the dimensions of the two previous groups. Additionally, the decay in the activity for the total inventory and the groups that conform it can be simulated by means of order 6 polynomials or smaller than describe its behavior appropriately. The results presented in this work, coupled

  18. Experimental determination of nuclear parameters for RP-0 reactor core; Determinacion experimental de los parametros nucleares para el nucleo tipo MTR del reactor nuclear RP-0

    Energy Technology Data Exchange (ETDEWEB)

    Cajacuri, Rafael A. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica

    2000-07-01

    In the nuclear reactor for investigations RP-0 which is in Lima, Peru, that is a open pool class reactor with 1 to 10 watts of power and as a nuclear fuel uranium 238 enriched to 20% constituted by elements of Material Testing Reactor fuel class. This has reflectors of graphite and moderator of water demineralized. In 1996/1997 was measured in this reactor the following parameters: position of the control bar that make critic the reactor, critic height of moderator, excess of reactivity of the nucleus, parameter of reactivity for vacuum, parameter of reactivity for temperature, reactivity of its control bar, levels of doses in the reactor. (author)

  19. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  20. Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test

    Directory of Open Access Journals (Sweden)

    Hyun-Sik Park

    2014-01-01

    Full Text Available To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests on the SBLOCA, CLOF, and PRHRS performances to support a verification of the SMART design and contribute to the SMART design licensing by providing proper test data for validating the system analysis codes. A typical scenario of SBLOCA was analyzed using the MARS-KS code to assess the thermal-hydraulic similarity between the SMART design and the VISTA-ITL facility, and a posttest simulation on a SBLOCA test for the shutdown cooling system line break has been performed with the MARS-KS code to assess its simulation capability for the SBLOCA scenario of the SMART design. The SBLOCA scenario in the SMART design was well reproduced using the VISTA-ITL facility, and the measured thermal-hydraulic data were properly simulated with the MARS-KS code.

  1. The Kallisti Limnes, carbon dioxide-accumulating subsea pools

    Science.gov (United States)

    2015-07-01

    Natural CO2 releases from shallow marine hydrothermal vents are assumed to mix into the water column, and not accumulate into stratified seafloor pools. We present newly discovered shallow subsea pools located within the Santorini volcanic caldera of the Southern Aegean Sea, Greece, that accumulate CO2 emissions from geologic reservoirs. This type of hydrothermal seafloor pool, containing highly concentrated CO2, provides direct evidence of shallow benthic CO2 accumulations originating from sub-seafloor releases. Samples taken from within these acidic pools are devoid of calcifying organisms, and channel structures among the pools indicate gravity driven flow, suggesting that seafloor release of CO2 at this site may preferentially impact benthic ecosystems. These naturally occurring seafloor pools may provide a diagnostic indicator of incipient volcanic activity and can serve as an analog for studying CO2 leakage and benthic accumulations from subsea carbon capture and storage sites.

  2. Application of Nondestructive Methods for Qualification of High Density Fuels in the IEA-R1 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, J.E.R.; Silva, A.T.; Domingos, D.B.; Terremoto, L.A.A. [Instituto de Pesquisas Energeticas e Nucleares, Comissao Nacional de Energia Nuclear (IPEN-CNEN/SP), Av.Prof. Lineu Prestes 2242, Cidade Universitaria 05508-000, Sao Paulo, SP (Brazil)

    2011-07-01

    The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralised water and having Beryllium and Graphite as reflectors. Since 1990, IPEN/CNEN-SP has been fabricating and qualifying its own U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}-Al dispersion fuels. The U{sub 3}O{sub 8}-Al dispersion fuel is qualified to a uranium density of 2.3 gU/cm{sup 3} and the U{sub 3}Si{sub 2}-Al dispersion fuel up to 3.0 gU/cm{sup 3}. The IEA-R1 reactor core is constituted of the fuels above, with low enrichment in U-235 (19.9% of U-235). Nowadays, IPEN/CNEN-SP is interested in qualifying the above dispersion fuels at higher densities. Fuel miniplates of U{sub 3}O{sub 8}-Al and U{sub 3}Si{sub 2}-Al fuels, with densities of 3.0 gU/cm{sup 3} and 4.8 gU/cm{sup 3}, respectively, which are the maximal uranium densities qualified worldwide for these dispersion fuels, were fabricated at IPEN/CNEN-SP. The miniplates were put in an irradiation device, with similar external dimensions of IEA-R1 fuel assemblies, which was placed in a peripheral position of the IEA-R1 reactor core. IPEN/CNEN-SP has no hot cells to provide destructive analysis of the irradiated fuel. As a consequence, non destructive methods are being used to evaluate irradiation performance of the fuel miniplates: i) monitoring the fuel miniplate performance during the IEA-R1 operation for the following parameters: reactor power, time of operation, neutron flux at the position of each fuel assembly, burnup, inlet and outlet water, and radiochemistry analysis of reactor water; ii) periodic underwater visual inspection of fuel miniplates and eventual sipping test for the fuel miniplate suspected of leakage. The miniplates are being periodically visually inspected by an underwater radiation-resistant camera inside the IEA-R1 reactor pool, to verify its integrity and its general plate surface conditions. A new special system was designed for the fuel miniplate swelling evaluation. The

  3. Effects of inoculum type and bulk dissolved oxygen concentration on achieving partial nitrification by entrapped-cell-based reactors.

    Science.gov (United States)

    Rongsayamanont, Chaiwat; Limpiyakorn, Tawan; Khan, Eakalak

    2014-07-01

    An entrapment of nitrifiers into gel matrix is employed as a tool to fulfill partial nitrification under non-limiting dissolved oxygen (DO) concentrations in bulk solutions. This study aims to clarify which of these two attributes, inoculum type and DO concentration in bulk solutions, is the decisive factor for partial nitrification in an entrapped-cell based system. Four polyvinyl alcohol entrapped inocula were prepared to have different proportions of nitrite-oxidizing bacteria (NOB) and nitrite-oxidizing activity. At a DO concentration of 3 mg l(-1), the number of active NOB cells in an inoculum was the decisive factor for partial nitrification enhancement. However, when the DO concentration was reduced to 2 mg l(-1), all entrapped cell inocula showed similar degrees of partial nitrification. The results suggested that with the lower bulk DO concentration, the preparation of entrapped cell inocula is not useful as the DO level becomes the decisive factor for achieving partial nitrification. Copyright © 2014 Elsevier Ltd. All rights reserved.

  4. Irradiation strategies for the production of Co{sup 60} in a MTR-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, Fabian E. [Investigacion Aplicada S.E. (INVAP), San Carlos de Bariloche (Argentina)

    1996-07-01

    There were analyzed some possible irradiation strategies for cobalt devices in a 10-MW MTR-type research, with radioisotope production criteria of 50000 Ci/year - provided by the extraction of pellets with 200 Ci/g as average specific activity. The present activity calculations rely on a series of six assumptions concerning the cycle length, the spatial treatment of pins with cobalt pellets and the bundle of pins, the calculation of absorption rates for each region and energy group the determination of appropriate macroscopic cross sections, the determination of appropriate fluxes, and the consideration of cobalt burnup in alternate cycles of T{sub 1} irradiation days followed by T{sub 2} decay days. It is shown the only irradiation strategy of two years of permanence of the cobalt device in different locations in the core and reflector for the reference bundle design - and some others strategies of three years of permanence - satisfying the design criteria. In addition, there studied alternative designs for the cobalt bundle, and the reactivity worth of cobalt for the safety analysis. Alternatives to the reference cobalt bundle seem to improve activities only in few percents. Typical uncertainties are estimated in a 10%. (author)

  5. Disinfection byproducts in swimming pool: occurrences, implications and future needs.

    Science.gov (United States)

    Chowdhury, Shakhawat; Alhooshani, Khalid; Karanfil, Tanju

    2014-04-15

    Disinfection of swimming pool water is essential to deactivate pathogenic microorganisms. Many swimming pools apply chlorine or bromine based disinfectants to prevent microbial growth. The chlorinated swimming pool water contains higher chlorine residual and is maintained at a higher temperature than a typical drinking water distribution system. It constitutes environments with high levels of disinfection by-products (DBPs) in water and air as a consequence of continuous disinfection and constant organic loading from the bathers. Exposure to those DBPs is inevitable for any bather or trainer, while such exposures can have elevated risks to human health. To date, over 70 peer-reviewed publications have reported various aspects of swimming pool, including types and quantities of DBPs, organic loads from bathers, factors affecting DBPs formation in swimming pool, human exposure and their potential risks. This paper aims to review the state of research on swimming pool including with the focus of DBPs in swimming pools, understand their types and variability, possible health effects and analyze the factors responsible for the formation of various DBPs in a swimming pool. The study identifies the current challenges and future research needs to minimize DBPs formation in a swimming pool and their consequent negative effects to bathers and trainers.

  6. Feasibility of Thermoelectric Waste Heat Recovery from Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byunghee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    A thermoelectric generator has the most competitive method to regenerate the waste heat from research reactors, because it has no limitation on operating temperature. In addition, since the TEG is a solid energy conversion device converting heat to electricity directly without moving parts, the regenerating power system becomes simple and highly reliable. In this regard, a waste heat recovery using thermoelectric generator (TEG) from 15-MW pool type research reactor is suggested and the feasibility is demonstrated. The producible power from waste heat is estimated with respect to the reactor parameters, and an application of the regenerated power is suggested by performing a safety analysis with the power. The producible power from TEG is estimated with respect to the LMTD of the HX and the required heat exchange area is also calculated. By increasing LMTD from 2 K to 20K, the efficiency and the power increases greatly. Also an application of the power regeneration system is suggested by performing a safety analysis with the system, and comparing the results with reference case without the power regeneration.

  7. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  8. Evolution of weld metals nanostructure and properties under irradiation and recovery annealing of VVER-type reactors

    Science.gov (United States)

    Gurovich, B.; Kuleshova, E.; Shtrombakh, Ya.; Fedotova, S.; Zabusov, O.; Prikhodko, K.; Zhurko, D.

    2013-03-01

    The results of VVER-440 steel Sv-10KhMFT and VVER-1000 steel SV-10KhGNMAA investigations by transmission electron microscopy, scanning electron microscopy, Auger-electron spectroscopy and mechanical tests are presented in this paper. The both types of weld metals with different content of impurities and alloying elements were studied after irradiations to fast neutron (E > 0.5 MeV) fluences in the wide range below and beyond the design values, after recovery annealing procedures and after re-irradiation following the annealing. The distinctive features of embrittlement kinetics of VVER-440 and VVER-1000 RPV weld metals conditioned by their chemical composition differences were investigated. It is shown that the main contribution into radiation strengthening within the design fluence can be attributed to radiation-induced precipitates, on reaching the design or beyond design values of fast neutron fluencies the main contribution into VVER-440 welds strengthening is made by radiation-induced dislocation loops, and in case of VVER-1000 welds - radiation-induced precipitates and grain-boundary phosphorous segregations. Recovery annealing of VVER-440 welds at 475 °C during 100 h causes irradiation-induced defects disappearance, transformation of copper enriched precipitates into bigger copper-rich precipitates with lower number density and leads to almost full recovery of mechanical properties followed by comparatively slow re-embrittlement rate. The recovery annealing temperature of VVER-1000 welds was higher - 565 °C during 100 h - to avoid temper brittleness. The annealing of VVER-1000 welds leads to almost full recovery of mechanical properties due to irradiation-induced defects disappearance and decrease in precipitates number density and grain-boundary segregation of phosphorus. The re-embrittlement rate of VVER-1000 weld during subsequent re-irradiation is at least not higher than the initial rate.

  9. Evaluation of Erosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brower, Jeffrey O. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Glazoff, Michael V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Eiden, Thomas J. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor; Rezvoi, Aleksey V. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Advanced Test Reactor

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, when the fuel elements were removed from the core and inspected, several thousand flow-assisted erosion pits and “horseshoeing” defects were readily observed on the surface of the several YA-type fuel elements (these are aluminum “dummy” plates that contain no fuel). In order to understand these erosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed scalloping

  10. Evaluation of Corrosion of the Dummy “EE” Plate 19 in YA Type ATR Fuel Element During Reactor PALM Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Brower, Jeffrey Owen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Glazoff, Michael Vasily [Idaho National Lab. (INL), Idaho Falls, ID (United States); Eiden, Thomas John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rezvoi, Aleksey Victor [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Advanced Test Reactor (ATR) Cycle 153B-1 was a 14-day, high-power, powered axial locator mechanism (PALM) operating cycle that completed on April 12, 2013. Cycle 153B-1 was a typical operating cycle for the ATR and did not result in any unusual plant transients. ATR was started up and shut down as scheduled. The PALM drive physically moves the selected experiments into and out of the core to simulate reactor startup and heat up, and shutdown and cooldown transients, while the reactor remains in steady state conditions. However, after the cycle was over, several thousand of the flow-assisted corrosion pits and “horseshoeing” defects were readily observable on the surface of the several YA-type fuel elements (these are “dummy” plates that contain no fuel). In order understand these corrosion phenomena a thermal-hydraulic model of coolant channel 20 on a YA-M fuel element was generated. The boundaries of the model were the aluminum EE plate of a YA-M fuel element and a beryllium reflector block with 13 horizontal saw cuts which represented regions of zero flow. The heat generated in fuel plates 1 through 18 was modeled to be passing through the aluminum EE plate. The coolant channel 20 width was set at 0.058 in. (58 mils). It was established that the horizontal saw cuts had a significant effect on the temperature of the coolant. The flow, which was expected to vary linearly with gradual heating of the coolant as it passed through the channel, was extremely turbulent. The temperature rise, which was expected to be a smooth “S” curve, was represented by a series temperature rise “humps,” which occurred at each horizontal saw cut in the beryllium reflector block. Each of the 13 saw cuts had a chamfered edge which resulted in the coolant flow being re-directed as a jet across the coolant channel into the surface of the EE plate, which explained the temperature rise and the observed sscalloping and possibly pitting degradation on the YA-M fuel elements. In

  11. Blood pressure and fasting lipid changes after 24 weeks’ treatment with vildagliptin: a pooled analysis in >2,000 previously drug-naïve patients with type 2 diabetes mellitus

    Directory of Open Access Journals (Sweden)

    Evans M

    2016-08-01

    Full Text Available Marc Evans,1 Anja Schweizer,2 James E Foley3 1Diabetes Resource Centre, Llandough Hospital, Cardiff, UK; 2Medical Affairs Cardio Metabolic, Novartis Pharma AG, Basel, Switzerland; 3Medical Affairs Cardio-Metabolic, Novartis Pharmaceuticals Corporation, East Hanover, NJ, USA Introduction: We have previously shown modest weight loss with vildagliptin treatment. Since body weight balance is associated with changes in blood pressure (BP and fasting lipids, we have assessed these parameters following vildagliptin treatment. Methods: Data were pooled from all double-blind, randomized, controlled, vildagliptin monotherapy trials on previously drug-naïve patients with type 2 diabetes mellitus who received vildagliptin 50 mg once daily (qd or twice daily (bid; n=2,108 and wherein BP and fasting lipid data were obtained. Results: Data from patients receiving vildagliptin 50 mg qd or bid showed reductions from baseline to week 24 in systolic BP (from 132.5±0.32 to 129.8±0.34 mmHg; P<0.0001, diastolic BP (from 81.2±0.18 to 79.6±0.19 mmHg; P<0.0001, fasting triglycerides (from 2.00±0.02 to 1.80±0.02 mmol/L; P<0.0001, very low density lipoprotein cholesterol (from 0.90±0.01 to 0.83±0.01 mmol/L; P<0.0001, and low density lipoprotein cholesterol (from 3.17±0.02 to 3.04±0.02 mmol/L; P<0.0001, whereas high density lipoprotein cholesterol increased (from 1.19±0.01 to 1.22±0.01 mmol/L; P<0.001. Weight decreased by 0.48±0.08 kg (P<0.001. Conclusion: This large pooled analysis demonstrated that vildagliptin shows a significant reduction in BP and a favorable fasting lipid profile that are associated with modest weight loss. Keywords: TG, HDL, LDL, body weight DPP-4 inhibitor, GLP-1 

  12. Modeling of condensation, stratification, and mixing phenomena in a pool of water

    Energy Technology Data Exchange (ETDEWEB)

    Li, H.; Kudinov, P.; Villanueva, W. (Royal Institute of Technology (KTH). Div. of Nuclear Power Safety, Stockholm (Sweden))

    2010-12-15

    This work pertains to the research program on Containment Thermal-Hydraulics at KTH. The objective is to evaluate and improve performance of methods, which are used to analyze thermal-hydraulics of steam suppression pools in a BWR plant under different abnormal transient and accident conditions. As a passive safety system, the function of steam pressure suppression pools is paramount to the containment performance. In the present work, the focus is on apparently-benign but intricate and potentially risk-significant scenarios in which thermal stratification could significantly impede the pool's pressure suppression capacity. For the case of small flow rates of steam influx, the steam condenses rapidly in the pool and the hot condensate rises in a narrow plume above the steam injection plane and spreads into a thin layer at the pool's free surface. When the steam flow rate increases significantly, momentum introduced by the steam injection and/or periodic expansion and shrink of large steam bubbles due to direct contact condensation can cause breakdown of the stratified layers and lead to mixing of the pool water. Accurate prediction of the pool thermal-hydraulics in such scenarios presents a computational challenge. Lumped-parameter models have no capability to predict temperature distribution of water pool during thermal stratification development. While high-order-accurate CFD (RANS, LES) methods are not practical due to excessive computing power needed to calculate 3D high-Rayleighnumber natural circulation flow in long transients. In the present work, a middleground approach is used, namely CFD-like model of the general purpose thermalhydraulic code GOTHIC. Each cell of 3D GOTHIC grid uses lumped parameter volume type closures for modeling of various heat and mass transfer processes at subgrid scale. We use GOTHIC to simulate POOLEX/PPOOLEX experiment, in order to (a) quantify errors due to GOTHIC's physical models and numerical schemes, and (b

  13. Mathematical model for evaluating the Krebs cycle flux with non-constant glutamate-pool size by 13C-NMR spectroscopy. Evidence for the existence of two types of Krebs cycles in cells.

    Science.gov (United States)

    Tran-Dinh, S; Beganton, F; Nguyen, T T; Bouet, F; Herve, M

    1996-12-01

    A practical method using matrix operations is proposed for studying the isotopic transformation of glutamate, or any other metabolite isotopomers, in the Krebs cycle. Two mathematical models were constructed for evaluating the Krebs cycle flux where the enrichment of [2-13C]acetyl-CoA is not 100% and the total glutamate concentration remains constant or varies during incubation. A comparative study of [1-13C]glucose metabolism was subsequently carried out using Saccharomyces cerevisiae cells from two different strains (ATCC-9763 and NCYC-239) by 13C-NMR spectroscopy and biochemical techniques. The results show that there are two types of Krebs cycles in cells. The first is represented by the ATCC cells which contain a small amount of 2-oxoglutarate dehydrogenase and hence the flux in the Krebs cycle is negligible. With [1-13C]glucose as a carbon source, the 13C-NMR spectra of glutamate exhibit the C2 and C4 resonances that are almost equivalent and much greater than that of the C3. Labeled metabolites derived from [1-13C]glucose enter the Krebs cycle at two points: oxaloacetate and citrate. The second cell type is represented by NCYC-239. The C2 and C3 areas are equivalent and smaller than the C4 resonance. The results suggest that labeled metabolites enter the Krebs cycle only at the citrate level via acetyl-CoA, 2-oxoglutarate dehydrogenase is present but pyruvate carboxylase is virtually absent or inactivated. When both are incubated with glucose, the total concentration of glutamate was found to decrease with the incubation time. The fraction of glutamate in isotopic exchange with the Krebs cycle in NCYC-239 cells is about 2.6% and the reduction in glutamate concentration is about 0.5%/min. Using our model, with a variable glutamate pool size, good agreement between the theoretical and experimental data is obtained.

  14. Irradiation Scheme Design of 14C Production on 49-2 Reactor

    Institute of Scientific and Technical Information of China (English)

    SUN; Zheng; LIU; Xing-min; XU; Zhi-long; ZHANG; Ya-dong

    2012-01-01

    <正>14C is a radioisotope of carbon, it is widely used in pharmacy, medical treatment, agriculture, reconnoiter and archaeology. 49-2 research reactor is a swimming pool style reactor which has operated for more than 40 years. The application of 49-2 reactor includes the radio nuclides production. Therefore, the technical scheme on 14C irradiation in 49-2 reactor should be prepared elaborately.

  15. Estimates of array and pool-construction variance for planning efficient DNA-pooling genome wide association studies

    Directory of Open Access Journals (Sweden)

    Earp Madalene A

    2011-11-01

    Full Text Available Abstract Background Until recently, genome-wide association studies (GWAS have been restricted to research groups with the budget necessary to genotype hundreds, if not thousands, of samples. Replacing individual genotyping with genotyping of DNA pools in Phase I of a GWAS has proven successful, and dramatically altered the financial feasibility of this approach. When conducting a pool-based GWAS, how well SNP allele frequency is estimated from a DNA pool will influence a study's power to detect associations. Here we address how to control the variance in allele frequency estimation when DNAs are pooled, and how to plan and conduct the most efficient well-powered pool-based GWAS. Methods By examining the variation in allele frequency estimation on SNP arrays between and within DNA pools we determine how array variance [var(earray] and pool-construction variance [var(econstruction] contribute to the total variance of allele frequency estimation. This information is useful in deciding whether replicate arrays or replicate pools are most useful in reducing variance. Our analysis is based on 27 DNA pools ranging in size from 74 to 446 individual samples, genotyped on a collective total of 128 Illumina beadarrays: 24 1M-Single, 32 1M-Duo, and 72 660-Quad. Results For all three Illumina SNP array types our estimates of var(earray were similar, between 3-4 × 10-4 for normalized data. Var(econstruction accounted for between 20-40% of pooling variance across 27 pools in normalized data. Conclusions We conclude that relative to var(earray, var(econstruction is of less importance in reducing the variance in allele frequency estimation from DNA pools; however, our data suggests that on average it may be more important than previously thought. We have prepared a simple online tool, PoolingPlanner (available at http://www.kchew.ca/PoolingPlanner/, which calculates the effective sample size (ESS of a DNA pool given a range of replicate array values. ESS can

  16. Morphology of drying blood pools

    Science.gov (United States)

    Laan, Nick; Smith, Fiona; Nicloux, Celine; Brutin, David; D-Blood project Collaboration

    2016-11-01

    Often blood pools are found on crime scenes providing information concerning the events and sequence of events that took place on the scene. However, there is a lack of knowledge concerning the drying dynamics of blood pools. This study focuses on the drying process of blood pools to determine what relevant information can be obtained for the forensic application. We recorded the drying process of blood pools with a camera and measured the weight. We found that the drying process can be separated into five different: coagulation, gelation, rim desiccation, centre desiccation, and final desiccation. Moreover, we found that the weight of the blood pool diminishes similarly and in a reproducible way for blood pools created in various conditions. In addition, we verify that the size of the blood pools is directly related to its volume and the wettability of the surface. Our study clearly shows that blood pools dry in a reproducible fashion. This preliminary work highlights the difficult task that represents blood pool analysis in forensic investigations, and how internal and external parameters influence its dynamics. We conclude that understanding the drying process dynamics would be advancement in timeline reconstitution of events. ANR funded project: D-Blood Project.

  17. Construction of an external electrode for determination of electrochemical corrosion potential in normal operational conditions of an BWR type reactor for hot cells; Construccion de un electrodo externo para determinacion del potencial de corrosion electroquimico en condiciones normales de operacion de un reactor tipo BWR para celdas calientes

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar T, J.A.; Rivera M, H.; Hernandez C, R. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    The behavior of the corrosion processes at high temperature requires of external devices that being capable to resist a temperature of 288 Centigrade and a pressure of 80 Kg/cm{sup 2}, to give stable and reproducible results of some variable and resisting physically and chemically the radiation. The external electrode of Ag/AgCl fulfils all the requirements in the determination of the electrochemical corrosion potential under normal operational conditions of a BWR type reactor in hot cells. (Author)

  18. Experimental investigations on turbulent mixing of hot upward flow and cold downward flow inside a chimney model of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sengupta, Samiran, E-mail: samiran_sengupta@yahoo.co.in [Research Reactor Design & Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Ghosh, Aniruddha [Research Reactor Design & Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, C. [Research Reactor Maintenance Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Vijayan, P.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai 400085 (India); Bhattacharya, S. [Research Reactor Design & Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sharma, R.C. [Reactor Group, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2016-02-15

    Highlights: • Simulated mixing of hot upward and cold downward flows in a chimney of a reactor. • Experiments in chimney model (2:9 scale) at Reynolds number (Re)—1.5 to 4.5 × 10{sup 5}. • Hot upward flow comes out of the chimney when bypass flow ratio (R) is zero. • Increase in ratio (R) reduces jet height, vortex spread height and temperature front height. • Effects of Re, chimney height and temperature differential are not significant. - Abstract: Experiments were conducted to study the turbulent mixing of hot upward flow and cold downward flow inside a scaled down model of chimney structure of a pool type nuclear research reactor. Open pool type nuclear reactors often use this type of chimney structures to prevent mixing of radioactive core outlet water directly into the reactor pool so that radiation field at the reactor pool top can be kept to a lower limit. The chimney structure is designed to facilitate guiding of the radioactive water towards the two outlet nozzles of the chimney and simultaneously allows drawing water from the reactor pool through the chimney top opening. The present work aims at studying flow mixing behaviour of hot and cold water inside a 2/9th scaled down model of the chimney structure experimentally. The ratio between the cold downward flow and the hot upward flow is varied between 0 and 0.15 to predict the extent of suppression of the hot upward flow within the chimney region for various bypass flow ratios. The Reynolds number of the hot upward flow considered in the experiment is about 1.5 × 10{sup 5} which corresponds to a flow rate of about 500 l min{sup −1}. The upward jet height and the temperature distribution were predicted from the experiment. It was observed that increase in bypass flow ratio reduces the upward jet height of hot water. Experiments were also carried out by increasing the flow rate to 1000 and 1500 l min{sup −1} corresponding to Reynolds numbers of 3 × 10{sup 5} and 4.5 × 10{sup 5

  19. Fault detection of sensors in nuclear reactors using self-organizing maps

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Paulo Roberto; Tiago, Graziela Marchi [Instituto Federal de Educacao, Ciencia e Tecnologia de Sao Paulo (IFSP), Sao Paulo, SP (Brazil); Bueno, Elaine Inacio [Instituto Federal de Educacao, Ciencia e Tecnologia de Sao Paulo (IFSP), Guarulhos, SP (Brazil); Pereira, Iraci Martinez, E-mail: martinez@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    In this work a Fault Detection System was developed based on the self-organizing maps methodology. This method was applied to the IEA-R1 research reactor at IPEN using a database generated by a theoretical model of the reactor. The IEA-R1 research reactor is a pool type reactor of 5 MW, cooled and moderated by light water, and uses graphite and beryllium as reflector. The theoretical model was developed using the Matlab Guide toolbox. The equations are based in the IEA-R1 mass and energy inventory balance and physical as well as operational aspects are taken into consideration. In order to test the model ability for fault detection, faults were artificially produced. As the value of the maximum calibration error for special thermocouples is +- 0.5 deg C, it had been inserted faults in the sensor signals with the purpose to produce the database considered in this work. The results show a high percentage of correct classification, encouraging the use of the technique for this type of industrial application. (author)

  20. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  1. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  2. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  3. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-05-27

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

  4. Calculation of absorbed doses to water pools in severe accident sequences

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F. [Oak Ridge National Lab., TN (United States)

    1991-12-01

    A methodology is presented for calculating the radiation dose to a water pool from the decay of uniformly distributed nuclides in that pool. Motivated by the need to accurately model radiolysis reactions of iodine, direct application is made to fission product sources dissolved or suspended in containment sumps or pools during a severe nuclear reactor accident. Two methods of calculating gamma absorption are discussed - one based on point-kernal integration and the other based on Monte Carlo techniques. Using least-squares minimization, the computed results are used to obtain a correlation that relates absorbed dose to source energy and surface-to-volume ratio of the pool. This correlation is applied to most relevant fission product nuclides and used to actually calculate transient sump dose rate in a pressurized-water reactor (PWR) severe accident sequence.

  5. Experimental investigation of particulate debris spreading in a pool

    Energy Technology Data Exchange (ETDEWEB)

    Konovalenko, A., E-mail: kono@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology (KTH) , Roslagstullsbacken 21, Stockholm 106 91 (Sweden); Basso, S., E-mail: simoneb@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology (KTH) , Roslagstullsbacken 21, Stockholm 106 91 (Sweden); Kudinov, P., E-mail: pkudinov@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology (KTH) , Roslagstullsbacken 21, Stockholm 106 91 (Sweden); Yakush, S.E., E-mail: yakush@ipmnet.ru [Institute for Problems in Mechanics of the Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow 119526 (Russian Federation)

    2016-02-15

    Termination of severe accident progression by core debris cooling in a deep pool of water under reactor vessel is considered in several designs of light water reactors. However, success of this accident mitigation strategy is contingent upon the effectiveness of heat removal by natural circulation from the debris bed. It is assumed that a porous bed will be formed in the pool in the process of core melt fragmentation and quenching. Debris bed coolability depends on its properties and system conditions. The properties of the bed, including its geometry are the outcomes of the debris bed formation process. Spreading of the debris particles in the pool by two-phase turbulent flows induced by the heat generated in the bed can affect the shape of the bed and thus influence its coolability. The goal of this work is to provide experimental data on spreading of solid particles in the pool by large-scale two-phase flow. The aim is to provide data necessary for understanding of separate effects and for development and validation of models and codes. Validated codes can be then used for prediction of debris bed formation under prototypic severe accident conditions. In PDS-P (Particulate Debris Spreading in the Pool) experiments, air injection at the bottom of the test section is employed as a means to create large-scale flow in the pool in isothermal conditions. The test section is a rectangular tank with a 2D slice geometry, it has fixed width (72 mm), adjustable length (up to 1.5 m) and allows water filling to the depth of up to 1 m. Variable pool length and depth allows studying two-phase circulating flows of different characteristic sizes and patterns. The average void fraction in the pool is determined by video recording and subsequent image processing. Particles are supplied from the top of the facility above the water surface. Results of several series of PDS-P experiments are reported in this paper. The influence of the gas flow rate, pool dimensions, particle density

  6. Development of Scaling Approach for Prediction of Terminal Spread Thickness of Melt Poured into a Pool of Water

    OpenAIRE

    Konovalenko, Alexander; Kudinov, Pavel

    2012-01-01

    Corium melt stabilization and long term cooling in a pool of water located beneath reactor vessel is adopted in several existing designs of light water reactors (LWRs) as an element in severe accident (SA) mitigation strategy. At certain conditions of melt release into the pool (e.g. large ratio of the vessel breach size to the pool depth), liquid melt can spread under water and reach a coolable configuration. Coolability of the melt is contingent on terminal spread thickness of the melt laye...

  7. Use of LEU in the aqueous homogeneous medical isotope production reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ball, R.M. [Babock & Wilcox, Lynchburg, VA (United States)

    1997-08-01

    The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.

  8. An overview of modeling methods for thermal mixing and stratification in large enclosures for reactor safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Per F. Peterson

    2010-10-01

    Thermal mixing and stratification phenomena play major roles in the safety of reactor systems with large enclosures, such as containment safety in current fleet of LWRs, long-term passive containment cooling in Gen III+ plants including AP-1000 and ESBWR, the cold and hot pool mixing in pool type sodium cooled fast reactor systems (SFR), and reactor cavity cooling system behavior in high temperature gas cooled reactors (HTGR), etc. Depending on the fidelity requirement and computational resources, 0-D steady state models (heat transfer correlations), 0-D lumped parameter based transient models, 1-D physical-based coarse grain models, and 3-D CFD models are available. Current major system analysis codes either have no models or only 0-D models for thermal stratification and mixing, which can only give highly approximate results for simple cases. While 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries. Due to prohibitive computational expenses for long transients in very large volumes, 3-D CFD simulations remain impractical for system analyses. For mixing in stably stratified large enclosures, UC Berkeley developed 1-D models basing on Zuber’s hierarchical two-tiered scaling analysis (HTTSA) method where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. This paper will present an overview on important thermal mixing and stratification phenomena in large enclosures for different reactors, major modeling methods and their advantages and limits, potential paths to improve simulation capability and reduce analysis uncertainty in this area for advanced reactor system analysis tools.

  9. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  10. Weir instability experiments in 1/4 reactor assembly model of PFBR

    Energy Technology Data Exchange (ETDEWEB)

    Thirumalai, M.; Gupta, P.K.; Anandaraj, M.; Prakash, V.; Vaidyanathan, G. [Indira Gandhi Centre for Atomic Research, Kalpakkam - 603102 (India)

    2005-07-01

    The construction of Prototype Fast Breeder Reactor (PFBR), a 500 MWe liquid sodium cooled reactor, has commenced at Kalpakkam in India. The main vessel of this pool type reactor acts as the primary containment in the reactor assembly. In order to keep the main vessel temperature below creep range and to reduce high temperature embrittlement, a small fraction of core flow (0.5 m3/s) is sent through an annular space formed between the main vessel and a cylindrical baffle (primary thermal baffle) to cool the vessel. The sodium after cooling the main vessel overflows the primary baffle (weir shell) and falls into another concentric pool of sodium separated from the cold pool by the secondary thermal baffle and then returns to the cold pool. These baffles, which are thin concentric shells, are prone to flow induced vibrations due to instability caused by sloshing and fluid-structure interaction. A similar vibration phenomenon was first observed during the commissioning of Super-Phenix reactor, which had got a similar main vessel cooling arrangement. In order to understand the phenomenon and also to provide necessary experimental back up to validate the analytical codes, weir instability experiments were conducted on a 1/4 scale stainless steel model installed in a water test loop. The experiments were conducted with flow rate and fall height as the varying parameters. The primary and secondary baffles in the model were instrumented with accelerometers and strain gages in circumferential and longitudinal directions at different locations to measure the vibration. At each fall height, the strain gage and accelerometer output signals were acquired and analyzed using an multichannel FFT analyzer. The baffle system became unstable under certain combinations of flow rate and fall height. From the analysis of shell vibration time plots, probability density functions, and spectra, the results showed that the instability of the weir shell was caused due to fluid structure

  11. Present state of the liner of the reactor; Estado actual del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F.; Raya A, R.; Mazon R, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-07-15

    When being presented to work the operation personnel of the reactor, on Monday January 10, 1983, they noticed that the reactor pool was overflowing of water and the floor of the room was partially flooded. The personnel proceeded to revise the feedwater systems to the pool, the Emergency Cooling System of the core and that of Water of Reinstatement, was found that the passing valve of this last it was lightly open. It was discovered that the water that was flooded in the floor of the room it came from the relief valves of the ports TW-1 and RW-2 and of three glides that were in the Thermal Column area. It was proceeded to lower the one level of water of the pool to their normal position and it was clean the water flooded in the salts. (Author)

  12. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-12-31

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  13. Reactor service life extension program

    Energy Technology Data Exchange (ETDEWEB)

    Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S.; Baumann, E.W.

    1990-01-01

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.

  14. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors; Validacion del sistema de codigos WIMS-SNAP para calculos en reactores nucleares tipo TRIGA-MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu

    2000-07-01

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  15. Cellular convection in vertical annuli of fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hemanath, M.G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)], E-mail: hemanath@igcar.gov.in; Meikandamurthy, C.; Ramakrishnan, V.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G. [Fast Reactor Technology Group, Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2007-08-15

    In the pool type fast reactors the roof structure is penetrated by a number of pumps and heat exchangers that are cylindrical in shape. Sandwiched between the free surface of sodium and the roof structure, is stagnant argon gas, which can flow in the annular space between the components and roof structure, as a thermosyphon. These thermosyphons not only transport heat from sodium to roof structure, but also result in cellular convection in vertical annuli resulting in circumferential temperature asymmetry of the penetrating components. There is need to know the temperature asymmetry as it can cause tilting of the components. Experiments were carried out in an annulus model to predict the circumferential temperature difference with and without sodium in the test vessel. Three-dimensional analysis was also carried out using PHOENICS CFD code and compared with the experiment. This paper describes the experimental details, the theoretical analysis and their comparison.

  16. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  17. Quality measure attainment with dapagliflozin plus metformin extended-release as initial combination therapy in patients with type 2 diabetes: a post hoc pooled analysis of two clinical studies

    Directory of Open Access Journals (Sweden)

    Bell KF

    2016-10-01

    Full Text Available Kelly F Bell, Arie Katz, John J Sheehan AstraZeneca, Wilmington, DE, USA Background: The use of quality measures attempts to improve safety and health outcomes and to reduce costs. In two Phase III trials in treatment-naive patients with type 2 diabetes, dapagliflozin 5 or 10 mg/d as initial combination therapy with metformin extended-release (XR significantly reduced glycated hemoglobin (A1C from baseline to 24 weeks and allowed higher proportions of patients to achieve A1C <7% vs dapagliflozin or metformin monotherapy. Objective: A pooled analysis of data from these two studies assessed the effect of dapagliflozin 5 or 10 mg/d plus metformin XR (combination therapy compared with placebo plus metformin XR (metformin monotherapy on diabetes quality measures. Quality measures include laboratory measures of A1C and low-density lipoprotein cholesterol (LDL-C as well as vital status measures of blood pressure (BP and body mass index (BMI. The proportion of patients achieving A1C, BP, and LDL-C individual and composite measures was assessed, as was the proportion with baseline BMI ≥25 kg/m2 who lost ≥4.5 kg. Subgroup analyses by baseline BMI were also performed. Results: A total of 194 and 211 patients were treated with dapagliflozin 5- or 10-mg/d combination therapy, respectively, and 409 with metformin monotherapy. Significantly higher proportions of patients achieved A1C ≤6.5%, <7%, or <8% with combination therapy vs metformin monotherapy (P<0.02. Significantly higher proportions of patients achieved BP <140/90 mmHg (P<0.02 for each dapagliflozin dose and BP <130/80 mmHg (P<0.02 with dapagliflozin 5 mg/d only with combination therapy vs metformin monotherapy. Similar proportions (29%–33% of patients had LDL-C <100 mg/dL across treatment groups. A higher proportion of patients with baseline BMI ≥25 kg/m2 lost ≥4.5 kg with combination therapy. Combination therapy had a more robust effect on patients with higher baseline BMI. Conclusion

  18. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  19. Rank Pooling for Action Recognition.

    Science.gov (United States)

    Fernando, Basura; Gavves, Efstratios; Oramas M, Jose Oramas; Ghodrati, Amir; Tuytelaars, Tinne

    2017-04-01

    We propose a function-based temporal pooling method that captures the latent structure of the video sequence data - e.g., how frame-level features evolve over time in a video. We show how the parameters of a function that has been fit to the video data can serve as a robust new video representation. As a specific example, we learn a pooling function via ranking machines. By learning to rank the frame-level features of a video in chronological order, we obtain a new representation that captures the video-wide temporal dynamics of a video, suitable for action recognition. Other than ranking functions, we explore different parametric models that could also explain the temporal changes in videos. The proposed functional pooling methods, and rank pooling in particular, is easy to interpret and implement, fast to compute and effective in recognizing a wide variety of actions. We evaluate our method on various benchmarks for generic action, fine-grained action and gesture recognition. Results show that rank pooling brings an absolute improvement of 7-10 average pooling baseline. At the same time, rank pooling is compatible with and complementary to several appearance and local motion based methods and features, such as improved trajectories and deep learning features.

  20. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  1. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T. (Inventor); Sahimi, Muhammad (Inventor); Fayyaz-Najafi, Babak (Inventor); Harale, Aadesh (Inventor); Park, Byoung-Gi (Inventor); Liu, Paul K. T. (Inventor)

    2011-01-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  2. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  3. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  4. 76 FR 11701 - Amendments to Commodity Pool Operator and Commodity Trading Advisor Regulations Resulting From...

    Science.gov (United States)

    2011-03-03

    ... reserves the right, but shall have no obligation, to review, pre-screen, filter, redact, refuse or remove... of the pool's assets that will be used to trade commodity interests, securities and other types of...) or from an investment of pool assets in investee pools or funds or other investments. * * * * * (l...

  5. Emission and transmission tomography systems to be developed for the future needs of Jules Horowitz material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kotiluoto, Petri [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)], E-mail: petri.kotiluoto@vtt.fi; Wasastjerna, Frej; Kekki, Tommi [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland); Sipilae, Heikki; Banzuzi, Kukka [Oxford Instruments Analytical Oy, Nihtisillankuja 5, P.O.Box 85, FI-02631 Espoo (Finland); Kinnunen, Petri; Heikinheimo, Liisa [VTT Technical Research Centre of Finland, P.O.Box 1000, FI-02044 VTT (Finland)

    2009-08-01

    The new 100 MW Jules Horowitz material testing reactor will be built in Cadarache, France. It will support, for instance, research on new types of innovative nuclear fuel. As a Finnish in-kind contribution, 3D emission and transmission tomography equipment will be delivered for both the reactor and the active component storage pool. The image reconstruction of activities inside the used nuclear fuel will be based on gamma spectrometry measurements. A new type of underwater digital X-ray linear detector array is under development for transmission imaging, based on GaAs and direct conversion of X-rays into an electrical signal. A shared collimator will be used for both emission and transmission measurements. Some preliminary design has been performed. For the current design, the expected gamma spectrometric response of a typical high-purity germanium detector has been simulated with MCNP for minimum and maximum source activities (specified by CEA) to be measured in future.

  6. Reactor Bolshoi Moshchnosti Kalani; Reacteurs RBMK

    Energy Technology Data Exchange (ETDEWEB)

    Bastien, D. [Conservatoire National des Arts et Metiers (CNAM), 75 - Paris (France)

    2000-01-01

    The Reactor Bolshoi Molshchnosti Kalani (RBMK) are pressure tubes reactor, boiling light water cooled. Exported since 1990 from the ex-USSR, they are today in three independent countries: Russian, Ukraine and Lithuania. Since this date, data exchange with the occident allowed the better knowledge of this reactor type. The design, the technical description (core, fuel, primary system), the safety and the improvement since Chernobyl are detailed. (A.L.B.)

  7. Chemical contaminants in swimming pools: Occurrence, implications and control.

    Science.gov (United States)

    Teo, Tiffany L L; Coleman, Heather M; Khan, Stuart J

    2015-03-01

    A range of trace chemical contaminants have been reported to occur in swimming pools. Current disinfection practices and monitoring of swimming pool water quality are aimed at preventing the spread of microbial infections and diseases. However, disinfection by-products (DBPs) are formed when the disinfectants used react with organic and inorganic matter in the pool. Additional chemicals may be present in swimming pools originating from anthropogenic sources (bodily excretions, lotions, cosmetics, etc.) or from the source water used where trace chemicals may already be present. DBPs have been the most widely investigated trace chemical contaminants, including trihalomethanes (THMs), haloacetic acids (HAAs), halobenzoquinones (HBQs), haloacetonitriles (HANs), halonitromethanes (HNMs), N-nitrosamines, nitrite, nitrates and chloramines. The presence and concentrations of these chemical contaminants are dependent upon several factors including the types of pools, types of disinfectants used, disinfectant dosages, bather loads, temperature and pH of swimming pool waters. Chemical constituents of personal care products (PCPs) such as parabens and ultraviolet (UV) filters from sunscreens have also been reported. By-products from reactions of these chemicals with disinfectants and UV irradiation have been reported and some may be more toxic than their parent compounds. There is evidence to suggest that exposure to some of these chemicals may lead to health risks. This paper provides a detailed review of various chemical contaminants reported in swimming pools. The concentrations of chemicals present in swimming pools may also provide an alternative indicator to swimming pool water quality, providing insights to contamination sources. Alternative treatment methods such as activated carbon filtration and advanced oxidation processes may be beneficial in improving swimming pool water quality.

  8. Interim irradiated fuel storage facility for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lolich, Jose [INVAP SE, Bariloche (Argentina)

    2002-07-01

    In most research reactors irradiated fuel discharged from the reactor is initially stored underwater inside the reactor building for along period of time. This allows for heat dissipation and fission product decay. In most cases this initial storage is done in a irradiated fuel storage facility pool located closed to the reactor core. After a certain cooling time, the fuel discharged should be relocated for long-term interim storage in a Irradiated Fuel Storage (IFS) Facility. IFS facilities are required for the safe storage of irradiated nuclear fuel before it is reprocessed or conditioned for disposal as radioactive waste. The IFS Facility described in this report is not an integral part of an operating nuclear reactor. This facility many be either co-located with nuclear facilities (such as a nuclear reactor or reprocessing plant) or sited independently of other nuclear facilities. (author)

  9. CO{sub 2} direct cycles suitable for AGR type reactors; Cycles directs de gaz carbonique applicables aux reacteurs du genre AGR

    Energy Technology Data Exchange (ETDEWEB)

    Maillet, E. [Commissariat a l' Energie Atomique. Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1967-10-01

    The perspectives given by the gas turbines under pressure, to build simple nuclear power plants and acieving significantly high yield, are specified. The CO{sub 2} is characterised by by good efficiency under moderate temperature (500 to 750 Celsius degrees), compactness and the simpleness of machines and the safe exploitation (supply, storage, relief cooling, thermosyphon). The revision of thermal properties of the CO{sub 2} and loss elements show that several direct cycles would fit in particular to the AGR type reactors. Cycles that would diverge a little from classical models and able to lead to power and heat generation can lead by simple means to the best results. Several satisfying solutions present for the starting up, the power regulation and the stopping. The nuclear power plant components and the functioning safety are equally considered in the present report. The conclusions stimulate the studies and realizations of carbon dioxide gas turbines in when approprite. [French] Les perspectives offertes par la turbine a gaz sous pression, pour construire des centrales nucleaires simples et de rendement progressivement eleve, se precisent actuellement. le CO{sub 2} se distingue par sa bonne efficacite a temperature moderee (500 a 750 degres celsius), la compacite et la simplicite des machines, et la surete qu'il apporte a l'exploitation ( approvisionnement, stockage, refroidissement de secours, thermosiphon). La revision des proprietes thermophysiques du CO{sub 2} et des elements de pertes montre que divers cycles directs conviendraient en particulier aux reacteurs agr ou derives. Des cycles s'ecartant peu des modeles classiques, et se pretant ulterieurement a la production simultanee d'electricite et de chaleur, peuvent conduire par des moyens simples aux meilleurs resultats d'ensemble. Plusieurs solutions satisfaisantes se presentent pour le demarrage, le reglage de la puissance et l'arret. Les composants de la centrale et la

  10. Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past Incidents, Experiments, and Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

    2016-04-17

    In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

  11. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  12. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suk, S.D.; Hahn, D. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2001-07-01

    Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

  13. Experimental studies of local coolant hydrodynamics using a scaled model of cassette-type fuel assembly of a KLT-40S reactor

    Science.gov (United States)

    Dmitriev, S. M.; Barinov, A. A.; Varentsov, A. V.; Doronkov, D. V.; Solntsev, D. N.; Khrobostov, A. E.

    2016-08-01

    The results of experimental studies of local hydrodynamic and mass exchange characteristics of the coolant flow behind the spacer grid in the fuel assembly of a KLT-40S reactor are presented. The experiments were aimed at the investigation of representative domains of the fuel assembly with three tracer injection regions. The studies were performed at the aerodynamic test facility using the tracer gas diffusion method. According to the theory of hydrodynamic similarity, the obtained experimental results can be transferred to full-scale coolant flow conditions in standard fuel assemblies. The analysis of the tracer concentration propagation made it possible to determine in detail the flow pattern and find the main regularities and specific features of the coolant flow behind the plate spacer grid of KLT-40S fuel assembly. The hydraulic resistance coefficient of the spacer grid was experimentally determined. The coefficients of mass exchange between cells for representative cells of the displacer region in the KLT-40S fuel assembly were calculated for the first time; these results are presented in the form of the "mixing matrix." The results of studies of local coolant flow hydrodynamics in the KLT-40S fuel assembly are used at AO Afrikantov OKBM for estimation of thermotechnical reliability of active cores for reactors of floating nuclear power stations. The experimental data on hydrodynamic and mass exchange characteristics are included in the database for verification of CDF codes and detailed cell-wise calculation of the active core for KLT-40S reactor installation. The results of these studies can be used at FSUE RFNC-VNIIEF for testing and verification of domestic three-dimensional hydrodynamic CFD codes ("Logos") that are applied for substantiation of newly designed reactor installations. Practical recommendations on the application of the obtained results in thermohydraulic calculations of the active core for the KLT-40S reactor will be worked out. Proposals

  14. Analysis of Microbial Communities in Biofilms from CSTR-Type Hollow Fiber Membrane Biofilm Reactors for Autotrophic Nitrification and Hydrogenotrophic Denitrification.

    Science.gov (United States)

    Shin, Jung-Hun; Kim, Byung-Chun; Choi, Okkyoung; Kim, Hyunook; Sang, Byoung-In

    2015-10-01

    Two hollow fiber membrane biofilm reactors (HF-MBfRs) were operated for autotrophic nitrification and hydrogenotrophic denitrification for over 300 days. Oxygen and hydrogen were supplied through the hollow fiber membrane for nitrification and denitrification, respectively. During the period, the nitrogen was removed with the efficiency of 82-97% for ammonium and 87-97% for nitrate and with the nitrogen removal load of 0.09-0.26 kg NH4(+)-N/m(3)/d and 0.10-0.21 kg NO3(-)-N/m(3)/d, depending on hydraulic retention time variation by the two HF-MBfRs for autotrophic nitrification and hydrogenotrophic denitrification, respectively. Biofilms were collected from diverse topological positions in the reactors, each at different nitrogen loading rates, and the microbial communities were analyzed with partial 16S rRNA gene sequences in denaturing gradient gel electrophoresis (DGGE). Detected DGGE band sequences in the reactors were correlated with nitrification or denitrification. The profile of the DGGE bands depended on the NH4(+) or NO3(-) loading rate, but it was hard to find a major strain affecting the nitrogen removal efficiency. Nitrospira-related phylum was detected in all biofilm samples from the nitrification reactors. Paracoccus sp. and Aquaspirillum sp., which are an autohydrogenotrophic bacterium and an oligotrophic denitrifier, respectively, were observed in the denitrification reactors. The distribution of microbial communities was relatively stable at different nitrogen loading rates, and DGGE analysis based on 16S rRNA (341f /534r) could successfully detect nitrate-oxidizing and hydrogen-oxidizing bacteria but not ammonium-oxidizing bacteria in the HF-MBfRs.

  15. Assessment of the implementation of a neutron measurement system during the commissioning of the Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sang Hoon; Suh, Sang Mun [Division of Research Reactor System Design, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cha, Han Ju [Dept. of Electrical Engineering, Intelligent Power Conversion Laboratory, Chungnam National University, Daejeon (Korea, Republic of)

    2017-04-15

    The Jordan Research and Training Reactor (JRTR) is the first research reactor in Jordan, the commissioning of which is ongoing. The reactor is a 5-MWth, open-pool type, light-water-moderated, and cooled reactor with a heavy water reflector system. The neutron measurement system (NMS) applied to the JRTR employs a wide-range fission chamber that can cover from source range to power range. A high-sensitivity boron trifluoride counter was added to obtain more accurate measurements of the neutron signals and to calibrate the log power signals; the NMS has a major role in the entire commissioning stage. However, few case studies exist concerning the application of the NMS to a research reactor. This study introduces the features of the NMS and the boron trifluoride counter in the JRTR and shares valuable experiences from lessons learned from the system installation to its early commissioning. In particular, the background noise relative to the signal-to-noise ratio and the NMS signal interlock are elaborated. The results of the count rates with the neutron source and the effects of the discriminator threshold are summarized.

  16. Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

    2011-02-15

    The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

  17. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Energy Technology Data Exchange (ETDEWEB)

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  18. Use of standard spectra for the short life radionuclides and ratios for long life radionuclides in the wastes of EDF PWR type reactors; Utilisation de spectres types pour les radionucleides a vie courte et de ratios pour les radionucleides a vie longue dans les dechets de REP EDF

    Energy Technology Data Exchange (ETDEWEB)

    Lantes, B. [Electricite de France (EDF-DPN/Groupe Environnement), 31 - Toulouse (France); Bienvenu, Ph. [CEA Cadarache, Dept. d' Etudes des Dechets, DED, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    This paper presents the type of declaration of radioactivity in the wastes of PWR type reactors park. Particularly, it insists on the justification of use of spectra for the declaration of short live radionuclides. It tackles the important developments of methods and measures of radiochemical analysis made by the Cea in order to determine the ratios to declare the long life radioisotopes. (N.C.)

  19. Grundfoss: Chlorination of Swimming Pools

    DEFF Research Database (Denmark)

    Hjorth, Poul G.; Hogan, John; Andreassen, Viggo

    1998-01-01

    Grundfos asked for a model, describing the problem of mixing chemicals, being dosed into water systems, to be developed. The application of the model should be dedicated to dosing aqueous solution of chlorine into swimming pools.......Grundfos asked for a model, describing the problem of mixing chemicals, being dosed into water systems, to be developed. The application of the model should be dedicated to dosing aqueous solution of chlorine into swimming pools....

  20. Grundfoss: Chlorination of Swimming Pools

    DEFF Research Database (Denmark)

    Hjorth, Poul G.; Hogan, John; Andreassen, Viggo

    1998-01-01

    Grundfos asked for a model, describing the problem of mixing chemicals, being dosed into water systems, to be developed. The application of the model should be dedicated to dosing aqueous solution of chlorine into swimming pools.......Grundfos asked for a model, describing the problem of mixing chemicals, being dosed into water systems, to be developed. The application of the model should be dedicated to dosing aqueous solution of chlorine into swimming pools....

  1. Scaledown of a methanol reactor

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.

    1983-07-01

    This article shows how it is possible to define operating conditions for pilot plants and development labs by scaling down a commercial reactor. Points out that scaledown consideration and experiment planning can be done in a similar manner for the boiling water-cooled, Lurgi-type reactor. Explains that although the design of large, single-train plants to produce methanol for fuel use has different economic objectives, product specifications, and technical constraints from the traditional commercial methanol plants, the same fundamental laws of thermodynamics and reaction kinetics apply to both types of operation.

  2. Stability analysis of a recycling circuit of a BWR type reactor. Theoretical study; Analisis de estabilidad de un circuito de recirculacion de un reactor del tipo BWR. Estudio teorico

    Energy Technology Data Exchange (ETDEWEB)

    Salinas H, J.G.; Espinosa P, G. [Universidad Autonoma Metropolitana-Iztapalapa, 09000 Mexico D.F. (Mexico); Gonzalez M, V.M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, 04000 Mexico D.F. (Mexico)

    2000-07-01

    The Technology, Regulation and Services Management of the National Commission of Nuclear Safety and Safeguards financed and in coordinate form with the I.P.H. Department of the Metropolitan Autonomous-Iztapalapa University developed the present project with the purpose of studying the effect of the recycling system on the linear stability of a BWR reactor whose reference central is the Laguna Verde power station. The present project forms part of a work series focused to the linear stability of the nuclear reactor of the Unit 1 at Laguna Verde power station. The components of the recycling system considered for the study of stability are the recycling external circuit (recycling pumps, valves) and the internal circuit (downcomer, jet pumps, lower full, driers, separators). The mathematical model is obtained applying mass balances and movement quantity in each one of the mentioned circuits. With respect to the nucleus model two regions are considered, the first one is made of a flow in one phase and the second one of a flow in two phases. For modelling the biphasic region it is considered homogenous flow. Generally it is studied the system behavior in the frequency domain starting from the transfer function applied to four operational states which correspond to the lower stability zone in the map power-flow of the Unit 1 of Laguna Verde power station. The Nyquist diagrams corresponding to each state as well as their characteristic frequency were determined. The results show that exists a very clear dependence of the power-flow relation on the stability of the system. It was found that the boiling length is an important parameter for the linear stability of the system. The obtained results show that the characteristic frequencies in unstability zones are similar to the reported data of the Unit 1 of the Laguna Verde power station in the event of power oscillations carried out in January 1995. (Author)

  3. Development of a neutronic model for the fuel of a high temperature gas reactor type PBMR; Desarrollo de un modelo neutronico para el combustible de un reactor de gas de alta temperatura tipo PBMR

    Energy Technology Data Exchange (ETDEWEB)

    Oropeza C, I.; Carmona H, R.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, Jiutepec, Morelos 62550 (Mexico)]. e-mail: ivonucci@prodigy.net.mx

    2008-07-01

    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 {sup b}enchmark{sup .} 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 {sup p}itch{sup 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. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  5. NASA Reactor Facility Hazards Summary. Volume 1

    Science.gov (United States)

    1959-01-01

    The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

  6. Fission product concentration evolution in sodium pool following a fuel subassembly failure in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Velusamy, K.; Selvaraj, P.; Kasinathan, N.; Chellapandi, P.; Chetal, S.; Bhoje, S. [Indira Gandhi Center for Atomic Research, Kalpakkam (India)

    2003-07-01

    During a fuel element failure in a liquid metal cooled fast breeder reactor, the fission products originating from the failed pins mix into the sodium pool. Delayed Neutron Detectors (DND) are provided in the sodium pool to detect such failures by way of detection of delayed neutrons emitted by the fission products. The transient evolution of fission product concentration is governed by the sodium flow distribution in the pool. Transient hydraulic analysis has been carried out using the CFD code PHOENICS to estimate fission product concentration evolution in hot pool. k- {epsilon} turbulence model and zero laminar diffusivity for the fission product concentration have been considered in the analysis. Times at which the failures of various fuel subassemblies (SA) are detected by the DND are obtained. It has been found that in order to effectively detect the failure of every fuel SA, a minimum of 8 DND in hot pool are essential.

  7. Nuclear reactor neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost