WorldWideScience

Sample records for treat reactor physics

  1. Physics design of the upgraded TREAT reactor

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Lell, R.M.; Liaw, J.R.; Ulrich, A.J.; Wade, D.C.; Yang, S.T.

    1980-01-01

    With the deferral of the Safety Test Facility (STF), the TREAT Upgrade (TU) reactor has assumed a lead role in the US LMFBR safety test program for the foreseeable future. The functional requirements on TU require a significant enhancement of the capability of the current TREAT reactor. A design of the TU reactor has been developed that modifies the central 11 x 11 fuel assembly array of the TREAT reactor such as to provide the increased source of hard spectrum neutrons necessary to meet the functional requirements. A safety consequence of the increased demands on TU is that the self limiting operation capability of TREAT has proved unattainable, and reliance on a safety grade Plant Protection System is necessary to ensure that no clad damage occurs under postulated low-probability reactivity accidents. With that constraint, the physics design of TU provides a means of meeting the functional requirements with a high degree of confidence

  2. 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.

  3. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    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

  4. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. 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 materials testing 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 2001 are summarised.

  5. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  6. The effect of carbon crystal structure on treat reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Swanson, R.W.; Harrison, L.J.

    1988-01-01

    The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory-West (ANL-W) is fueled with urania in a graphite and carbon mixture. This fuel was fabricated from a mixture of graphite flour, thermax (a thermatomic carbon produced by ''cracking'' natural gas), coal-tar resin and U/sub 3/O/sub 8/. During the fabrication process, the fuel was baked to dissociate the resin, but the high temperature necessary to graphitize the carbon in the thermax and in the resin was avoided. Therefore, the carbon crystal structure is a complex mixture of graphite particles in a nongraphitized elemental carbon matrix. Results of calculations using macroscopic carbon cross sections obtained by mixing bound-kernel graphite cross sections for the graphitized carbon and free-gas carbon cross sections for the remainder of the carbon and calculations using only bound-kernel graphite cross sections are compared to experimental data. It is shown that the use of the hybridized cross sections which reflect the allotropic mixture of the carbon in the TREAT fuel results in a significant improvement in the accuracy of calculated neutronics parameters for the TREAT reactor. 6 refs., 2 figs., 3 tabs.

  7. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  8. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and 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 on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research 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. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  9. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and 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 on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research 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. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  10. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  11. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  12. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    Barjon, Robert

    1975-01-01

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude [fr

  13. Reactor physics of CANFLEX

    International Nuclear Information System (INIS)

    Sim, K. S.; Min, Byung Joo.

    1997-07-01

    Characteristic of reactor physics for CANFLEX-NU fuel core were calculated using final fuel design data. The results of analysis showed that there was no impact on reactor operations and safety. The above results of calculations and analysis were described in the physics design for CANFLEX-NU core. Various fuel models were evaluated for selecting high burnup fuel using recovered uranium. It is judged to be worse effects for reactor safety. Hence, the use of graphite within fuel was proposed and its results showed to be better. The analysis system of reactor physics for design and analysis of high burnup fuel was evaluated. Lattice codes and core code were reviewed. From the results, the probability of WIMS-AECL and HELIOS is known to be high for analysis of high burnup fuel. For the core code, RFSP, it was evaluated that the simplified 2 group equation should be replaced by explicit 2 group equation. This report also describes about the status of critical assemblies in other countries. (author). 58 refs., 41 tabs., 126 figs

  14. Reactor physics computations

    International Nuclear Information System (INIS)

    Shapiro, A.

    1977-01-01

    Those reactor-core calculations which provide the effective multiplication factor (or eigenvalue) and the stationary (or fundamental mode) neutron-flux distribution at selected times during the lifetime of the core are considered. The multiplication factor is required to establish the nuclear composition and configuration which satisfy criticality and control requirements. The steady-state flux distribution must be known to calculate reaction rates and power distributions which are needed for the thermal, mechanical and shielding design of the reactor, as well as for evaluating refueling requirements. The calculational methods and techniques used for evaluating the nuclear design information vary with the type of reactor and with the preferences and prejudices of the reactor-physics group responsible for the calculation. Additionally, new methods and techniques are continually being developed and made operational. This results in a rather large conglomeration of methods and computer codes which are available for reactor analysis. The author provides the basic calculational framework and discusses the more prominent techniques which have evolved. (Auth.)

  15. Studies on reactor physics

    International Nuclear Information System (INIS)

    1960-01-01

    Most of the peaceful applications of atomic energy are inherently dependent on advances in the science and technology of nuclear reactors, and aspects of this development are part of a major programme of the International Atomic Energy Agency. The most useful role that the Agency can play is as a co-ordinating body or central forum where the trends can be reviewed and the results assessed. Some of the basic studies are carried out by members of the Agency's own scientific staff. The Agency also convenes groups of experts from different countries to examine a particular problem in detail and make any necessary recommendations. Some of the important subjects are discussed at international scientific meetings held by the Agency. One of the subjects covered by such studies is the physics of nuclear reactors and a specific topic recently discussed was Codes for Reactor Computations, on which a seminar was held in Vienna in April this year. Another The members of the Panel described the development of heavy water reactors, the equipment and methods of research currently used, and plans for further development in their respective countries meeting of Panel of Experts on Heavy Water Lattices was held in Vienna in August 1959

  16. Reactor physics for non-nuclear engineers

    International Nuclear Information System (INIS)

    Lewis, E.E.

    2011-01-01

    A one-term undergraduate course in reactor physics is described. The instructional format is strongly influenced by its intended audience of non-nuclear engineering students. In contrast to legacy treatments of the subject, the course focuses on the physics of nuclear power reactors with no attempt to include instruction in numerical methods. The multi-physics of power reactors is emphasized highlighting the close interactions between neutronic and thermal phenomena in design and analysis. Consequently, the material's sequencing also differs from traditional treatments, for example treating kinetics before the neutron diffusion is introduced. (author)

  17. Standards for reference reactor physics measurements

    International Nuclear Information System (INIS)

    Harris, D.R.; Cokinos, D.M.; Uotinen, V.

    1990-01-01

    Reactor physics analysis methods require experimental testing and confirmation over the range of practical reactor configurations and states. This range is somewhat limited by practical fuel types such as actinide oxides or carbides enclosed in metal cladding. On the other hand, this range continues to broaden because of the trend of using higher enrichment, if only slightly enriched, electric utility fuel. The need for experimental testing of the reactor physics analysis methods arises in part because of the continual broadening of the range of core designs, and in part because of the nature of the analysis methods. Reactor physics analyses are directed primarily at the determination of core reactivities and reaction rates, the former largely for reasons of reactor control, and the latter largely to ensure that material limitations are not violated. Errors in these analyses can be regarded as being from numerics, from the data base, and from human factors. For numerical, data base, and human factor reasons, then, it is prudent and customary to qualify reactor physical analysis methods against experiments. These experiments can be treated as being at low power or at high power, and each of these types is subject to an American National Standards Institute standard. The purpose of these standards is to aid in improving and maintaining adequate quality in reactor physics methods, and it is from this point of view that the standards are examined here

  18. Nuclear reactors: physics and materials

    Energy Technology Data Exchange (ETDEWEB)

    Yadigaroglu, G

    2005-07-01

    In the form of a tutorial addressed to non-specialists, the article provides an introduction to nuclear reactor technology and more specifically to Light Water Reactors (LWR); it also shows where materials and chemistry problems are encountered in reactor technology. The basics of reactor physics are reviewed, as well as the various strategies in reactor design and the corresponding choices of materials (fuel, coolant, structural materials, etc.). A brief description of the various types of commercial power reactors follows. The design of LWRs is discussed in greater detail; the properties of light water as coolant and moderator are put in perspective. The physicochemical and metallurgical properties of the materials impose thermal limits that determine the performance and the maximum power a reactor can deliver. (author)

  19. Physical security at research reactors

    International Nuclear Information System (INIS)

    Clark, R.A.

    1977-01-01

    Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

  20. Reactors physics. Bases of nuclear physics

    International Nuclear Information System (INIS)

    Diop, Ch.M.

    2006-01-01

    The aim of nuclear reactor physics is to quantify the relevant macroscopic data for the characterization of the neutronic state of a reactor core and to evaluate the effects of radiations (neutrons and gamma radiations) on organic matter and on inorganic materials. This first article presents the bases of nuclear physics in the context of nuclear reactors: 1 - reactor physics and nuclear physics; 2 - atomic nucleus - basic definitions: nucleus constituents, dimensions and mass of the atomic nucleus, mass defect, binding energy and stability of the nucleus, strong interaction, nuclear momentums of nucleons and nucleus; 3 - nucleus stability and radioactivity: equation of evolution with time - radioactive decay law; alpha decay, stability limit of spontaneous fission, beta decay, electronic capture, gamma emission, internal conversion, radioactivity, two-body problem and notion of radioactive equilibrium. (J.S.)

  1. Graphite reactor physics

    International Nuclear Information System (INIS)

    Bacher, P.; Cogne, F.

    1964-01-01

    The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm 2 , channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors) [fr

  2. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

  3. Physics of pressurized water reactors

    International Nuclear Information System (INIS)

    Gruen, A.

    1980-01-01

    The objective of this lecture is to demonstrate typical problems and solutions encountered in the design and operation of PWR power plants. The examples selected for illustration refer to PWR's of KWU design and to results of KWU design methods. In order to understand the physics of a power reactor it is necessary to have some knowledge of the structure and design of the power plant system of which the reactor is a part. It is therefore assumed that the reader is familiar with the design of the more important components and systems of a PWR, such as fuel assemblies, control assemblies, core lay-out, reactor coolant system, instrumentation. (author)

  4. Fundamentals of Nuclear Reactor Physics

    CERN Document Server

    Lewis, E E

    2008-01-01

    This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

  5. Physical experiments. Reactor theory

    International Nuclear Information System (INIS)

    Korn, H.; Werle, H.; Bluhm, H.; Fieg, G.; Kappler, F.; Kuhn, D.; Lalovic, M.; Woll, D.; Kuefner, K.; Woznicki, Z.; Buckel, G.; Stehle, B.; Borgwaldt, H.

    1975-01-01

    The γ-spectrum in SNEAK 9C-1 and 9C-2 was measured by means of Si(Li) solid state detectors for verification of methods of shielding calculation. The blanket spectra turned out to be slightly harder than the spectra in the fissile zone; the plutonium spectra are slightly harder than the respective uranium spectra. This result is expected to be explained by studies to be carried out on the basis of a γ-transport program. For reactor theoretical calculations two 2-dimensional diffusion programs were compared with each other, and a 3-dimensional diffusion program was compared with a flux synthesis program. An improved source iteration scheme was drafted for the Karlsruhe Monte Carlo code. (orig.) [de

  6. HTR characteristics affecting reactor physics

    International Nuclear Information System (INIS)

    Ehlers, K.

    1980-01-01

    A physical description of high-temperature has-cooled reactors is given, followed by an overview of HTR characteristics. The emphasis is placed on the HTR fuel cycle alternatives and thermohydraulics of pebble bed core. Some prospects of HTRs in the Federal Republic of Germany are also presented

  7. Multimedia on nuclear reactors physics

    International Nuclear Information System (INIS)

    Dies, Javier; Puig, Francesc

    2010-01-01

    The paper present an example of measures that have been found to be effective in the development of innovative educational and training technology. A multimedia course on nuclear reactor physics is presented. This material has been used for courses at master level at the universities; training for engineers at nuclear power plant as modular 2 weeks course; and training operators of nuclear power plant. The multimedia has about 785 slides and the text is in English, Spanish and French. (authors)

  8. Physical protection of power reactors

    International Nuclear Information System (INIS)

    Darby, J.L.

    1979-01-01

    Sandia Laboratories has applied a systematic approach to designing physical protection systems for nuclear facilities to commercial light-water reactor power plants. A number of candidate physical protection systems were developed and evaluated. Focus is placed on the design of access control subsystems at each of three plant layers: the protected area perimeter, building surfaces, and vital areas. Access control refers to barriers, detectors, and entry control devices and procedures used to keep unauthorized personnel and contraband out of the plant, and to control authorized entry into vital areas within the plant

  9. Reactor physics using a microcomputer

    International Nuclear Information System (INIS)

    Murray, R.L.

    1983-01-01

    The object of the work reported is to develop educational computer modules for all aspects of reactor physics. The modules consist of a description of the theory, mathematical method, computer program listing, sample calculations, and problems for the student, along with a card deck. Modules were first written in FORTRAN for an IBM 360/75, then later in BASIC for microcomputers. Problems include: limitation of equipment, choice of format for the program, the variety of dialects of BASIC used in the different microcomputer and peripherals brands, and knowing when to quit in the process of developing a program

  10. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  11. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

    2008-01-01

    As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

  12. Photocatalytic reactors for treating water pollution with solar illumination. I: a simplified analysis for batch reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Inst. fuer Technische Chemie, Univ. Hannover, Hannover (Germany); Brandi, R.J.; Cassano, A.E. [INTEC (Univ. Nacional del Litoral and CONICET), Santa Fe (Argentina)

    2003-07-01

    Usual applications of photocatalytic reactors for treating wastewater exhibit the difficulty of handling fluids having varying composition and/or concentrations; thus, a detailed kinetic representation may not be possible. When the catalyst activation is obtained employing solar illumination an additional complexity always coexists: solar fluxes are permanently changing with time. For comparing different reacting systems under similar operating conditions and to provide approximate estimations for scaling up purposes, simplified models may be useful. For these approximations the model parameters should be restricted as much as possible to initial physical and boundary conditions such as: initial concentrations (expressed as such or as TOC measurements), flow rate or reactor volume, irradiated reactor area, incident radiation fluxes and a fairly simple experimental observation such as the photonic efficiency. A combination of a new concept: the ''actual observed photonic efficiency'' with ideal reactor models and empirical kinetic rate expressions can be used to provide rather simple working equations that can be efficiently used to describe the performance of practical reactors. In this paper, the method has been developed for the case of a photocatalytic batch reactor (PBR). (orig.)

  13. Physical-chemical and operational performance of an anaerobic baffled reactor (ABR treating swine wastewater - 10.4025/actascitechnol.v32i4.7203

    Directory of Open Access Journals (Sweden)

    Erlon Lopes Pereira

    2010-12-01

    Full Text Available Since hog raising concentrates a huge amount of swine manure in small areas, it is considered by the environmental government organizations to be one of the most potentially pollutant activities. Therefore the main objective of this research was to evaluate by operational criteria and removal efficiency, the performance of a Anaerobic Baffled Reactor (ABR, working as a biological pre-treatment of swine culture effluents. The physical-chemical analyses carried out were: total COD, BOD5, total solids (TS, fix (TFS and volatiles (TVS, temperature, pH, total Kjeldahl nitrogen, phosphorus, total acidity and alkalinity. The ABR unit worked with an average efficiency of 65.2 and 76.2%, respectively, concerning total COD and BOD5, with a hydraulic retention time (HRT about 15 hours. The results for volumetric organic loading rate (VOLR, organic loading rate (OLR and hydraulic loading rate (HLR were: 4.46 kg BOD m-3 day-1; 1.81 kg BOD5 kg TVS-1 day-1 and 1.57 m3 m-3 day-1, respectively. The average efficiency of the whole treatment system for total COD and BOD5 removal were 66.5 and 77.8%, showing an adequate performance in removing the organic matter from swine wastewater.

  14. The mechanics in the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.

    1998-01-01

    This meeting of the 24 november 1998, took place in Paris and was organized by the SFEN. After three plenary sessions a technical meeting dealt on the mechanics in reactors physics. The plenary papers presented the state of the art in the PWR type reactors and fast neutron reactors systems and in the thermonuclear reactors system. Five more technical papers presented the seismic behavior of the reactors cores, the fuel-cladding interactions, the defects harmfulness in the fracture mechanics and the fuel rods control system wear. (A.L.B.)

  15. Advances in reactor physics education: Visualization of reactor parameters

    International Nuclear Information System (INIS)

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-01-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  16. Impacts on power reactor health physics programs

    International Nuclear Information System (INIS)

    Meyer, B.A.

    1991-01-01

    The impacts on power reactor health physics programs form implementing the revised 10 CFR Part 20 will be extensive and costly. Every policy, program, procedure and training lesson plan involving health physics will require changes and the subsequent retraining of personnel. At each power reactor facility, hundreds of procedures and thousands of people will be affected by these changes. Every area of a power reactor health physics program will be affected. These areas include; ALARA, Respiratory Protection, Exposure Control, Job Coverage, Dosimetry, Radwaste, Effluent Accountability, Emergency Planning and Radiation Worker Training. This paper presents how power reactor facilities will go about making these changes and gives possible examples of some of these changes and their impact on each area of power reactor health physics program

  17. The physics of nuclear reactors

    CERN Document Server

    Marguet, Serge

    2017-01-01

    This comprehensive volume offers readers a progressive and highly detailed introduction to the complex behavior of neutrons in general, and in the context of nuclear power generation. A compendium and handbook for nuclear engineers, a source of teaching material for academic lecturers as well as a graduate text for advanced students and other non-experts wishing to enter this field, it is based on the author’s teaching and research experience and his recognized expertise in nuclear safety. After recapping a number of points in nuclear physics, placing the theoretical notions in their historical context, the book successively reveals the latest quantitative theories concerning: •   The slowing-down of neutrons in matter •   The charged particles and electromagnetic rays •   The calculation scheme, especially the simplification hypothesis •   The concept of criticality based on chain reactions •   The theory of homogeneous and heterogeneous reactors •   The problem of self-shielding �...

  18. Physical exercise in treating obesity

    Directory of Open Access Journals (Sweden)

    Victor Keihan Rodrigues Matsudo

    2006-03-01

    Full Text Available Undoubtedly, no regular practice of physical exercise is one of thefactors that determine the global epidemics of weight excess andobesity in all age groups. Taking up physical activities regularlysince the initial stages of life (childhood, during adolescence andmaintaining them in adulthood – from young adults to over 50 yearsof age - is essential to assure an appropriate control of weight andbody fat. The general recommendation of physical exercise for goodhealth is to practice at least 30 minutes of moderate activities, atleast five days a week, and preferably every day. When the purposeis to lose and control weight in overweighed and obese individuals,the minimum practice should last 60 minutes/day, preferably 90minutes/day, at least five days/week, in a continuous or accumulatedmanner. Physical exercise is associated with several physical,psychological and social benefits that justify it inclusion as a crucialstrategy to prevent and treat overweight and obesity in any agegroup. Apart from moderate aerobic physical exercise, such aswalking, cycling, swimming, or more vigorous activities, such asjogging or running, resistance exercises and changes in lifestyle areessential, together with re-education of eating habits, to fight theepidemics of overweight and obesity. Besides the effect of weightcontrol, reduced body fat, prevention of weight gain and maintenanceof lean mass, physical exercise is related to a better lipid profile andreduced risk of associated diseases, such as diabetes, hypertension,metabolic syndrome, cardiovascular diseases and, consequently,lower risk of death.

  19. General remarks on fast neutron reactor physics

    International Nuclear Information System (INIS)

    Barre, J.Y.

    1980-01-01

    The main aspects of fast reactor physics, presented in these lecture notes, are restricted to LMFBR's. The emphasis is placed on the core neutronic balance and the burn-up problems. After a brief description of the power reactor main components and of the fast reactor chronology, the fundamental parameters of the one-group neutronic balance are briefly reviewed. Then the neutronic burn-up problems related to the Pu production and to the doubling time are considered

  20. OKLO: Fossil nuclear reactors. Physical study

    International Nuclear Information System (INIS)

    Naudet, R.

    1991-04-01

    This book presents a study of Oklo reactors, based essentially on physics and particularly neutronics but reviewing also all what is known on this topic, regrouping observations, measurement results and interpretative calculations. A remarkable characteristic of the study is the use of sophisticated reactor calculation methods for analysis of what happened two billion years ago in a uranium deposit. 200 refs [fr

  1. Technical specifications: Health Physics Research Reactor

    International Nuclear Information System (INIS)

    1986-03-01

    These technical specifications define the key limitations that must be observed for safe operation of the Health Physics Research Reactor (HPRR) and an envelope of operation within which there is assurance that these limits will not be exceeded

  2. Reactor physics activities in NEA member countries

    International Nuclear Information System (INIS)

    1990-01-01

    This document is a compilation of National activity reports presented at the thirty-third Meeting of the NEA Committee on Reactor Physics, held at OECD Headquarters, Paris, from 15th - 19th October 1990

  3. Review of the treat upgrade reactor scram system reliability analysis

    International Nuclear Information System (INIS)

    Montague, D.F.; Fussell, J.B.; Krois, P.A.; Morelock, T.C.; Knee, H.E.; Manning, J.J.; Haas, P.M.; West, K.W.

    1984-10-01

    In order to resolve some key LMFBR safety issues, ANL personnel are modifying the TREAT reactor to handle much larger experiments. As a result of these modifications, the upgraded Treat reactor will not always operate in a self-limited mode. During certain experiments in the upgraded TREAT reactor, it is possible that the fuel could be damaged by overheating if, once the computer systems fail, the reactor scram system (RSS) fails on demand. To help ensure that the upgraded TREAT reactor is shut down when required, ANL personnel have designed a triply redundant RSS for the facility. The RSS is designed to meet three reliability goals: (1) a loss of capability failure probability of 10 -9 /demand (independent failures only); (2) an inadvertent shutdown probability of 10 -3 /experiment; and (3) protection agaist any known potential common cause failures. According to ANL's reliability analysis of the RSS, this system substantially meets these goals

  4. Physics and safety of advanced research reactors

    International Nuclear Information System (INIS)

    Boening, K.; Hardt, P. von der

    1987-01-01

    Advanced research reactor concepts are presently being developed in order to meet the neutron-based research needs of the nineties. Among these research reactors, which are characterized by an average power density of 1-10 MW per liter, highest priority is now generally given to the 'beam tube reactors'. These provide very high values of the thermal neutron flux (10 14 -10 16 cm -2 s -1 ) in a large volume outside of the reactor core, which can be used for sample irradiations and, in particular, for neutron scattering experiments. The paper first discusses the 'inverse flux trap concept' and the main physical aspects of the design and optimization of beam tube reactors. After that two examples of advanced research reactor projects are described which may be considered as two opposite extremes with respect to the physical optimization principle just mentioned. The present situation concerning cross section libraries and neutronic computer codes is more or less satisfactory. The safety analyses of advanced research reactors can largely be updated from those of current new designs, partially taking advantage of the immense volume of work done for power reactors. The paper indicates a few areas where generic problems for advanced research reactor safety are to be solved. (orig.)

  5. Activity report of Reactor Physics Section - 1985

    International Nuclear Information System (INIS)

    John, T.M.

    1986-01-01

    This Activity Report contains brief summaries of different studies made in Reactor Physics Section during the year 1985. These are presented under the headings Nuclear Data Processing and Validation, Reactor Design and Analysis, Safety and Noise Analysis, Radiation Transport and Shielding, Reactor Physics Experiments and Statistical Physics. The work on nuclear data during this period comprises primarily of validation of data of 232 Th and 233 U as a part of participation in the Co-ordinated Research Programme (CRP) under IAEA research contract. The most significant event during 1985 at this centre has been the first criticality of FBTR (Fast Breeder Test Reactor), which was achieved on the 18th of October. Reactor Physics Section has played a key role in this event by carrying out the first approach to criticality with fuel loading in a safe manner and conducting some low power reactor physics experiments which are discussed. The studies made in the field reactor safety and shielding are also connected mainly with the FBTR problems in addition to some work on the PFBR (Prototype Fast Breeder Reactor) detailed design of which has been just started. Studies pertaining to the other two Co-ordinated Research Programmes (CRP) under IAEA contract, namely (1) on the comparative assessment of processing techniques for the analysis of sodium boiling noise detection and, (2) on the contribution of advanced reactors to energy supply have been continued during this year. At the end of this report, a list of publications made by the members of the section and also the sectional seminars held during this period is included. (author)

  6. Space reactor fuel element testing in upgraded TREAT

    International Nuclear Information System (INIS)

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

  7. Space reactor fuel element testing in upgraded TREAT

    Science.gov (United States)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

  8. Nuclear reactor safety: physics and engineering aspects

    International Nuclear Information System (INIS)

    Kinchin, G.H.

    1982-01-01

    In order to carry out the sort of probabilistic analysis referred to by Farmer (Contemp. Phys.; 22:349(1981)), it is necessary to have a good understanding of the processes involved in both normal and accident conditions in a nuclear reactor. Some of these processes, for a variety of different reactor systems, are considered in sections dealing with the neutron chain reaction, the removal of heat from the reactor, material problems, reliability of protective systems and a number of specific topics of particular interest from the point of view of physics or engineering. (author)

  9. Physics: A New Reactor Physics Analysis Toolkit

    International Nuclear Information System (INIS)

    Rabiti, C.; Wang, Y.; Palmiotti, G.; Hiruta, H.; Cogliati, J.; Alfonsi, A.

    2011-01-01

    In the last year INL has internally pursued the development of a new reactor analysis tool: PHISICS. The software is built in a modular approach to simplify the independent development of modules by different teams and future maintenance. Most of the modules at the time of this summary are still under development (time dependent transport driver, depletion, cross section I/O and interpolation, generalized perturbation theory), while the transport solver INSTANT (Intelligent Nodal and Semi-structured Treatment for Advanced Neutron Transport) has already been widely used1, 2, 3, 4. For this reason we will focus mainly on the presentation of the transport solver INSTANT

  10. Reactor physics experiment plan using TCA

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors, which aims at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. This report is to plan critical experiments using TCA in JAERI. Critical Experiments performed so far in Europe and Japan are reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX-fuel rods used in the experiments is obtained by calculations and modification of the equipment for the experiments are shown. New MOX fuel and UO{sub 2} fuel rods are necessary for the RMWR critical experiments. Number of MOX fuel rods is 1000 for Plutonium fissile enrichment of 5 wt%, 1000 for 10 wt%, 1500 for 15 wt% and 500 for 20 wt%, respectively. Depleted UO{sub 2} fuel rods for blanket/buffer region are 4000. Driver fuel rods of 4.9 wt% UO{sub 2} are 3000. Modification of TCA facility is requested to treat the large amount of MOX fuels from safety point of view. Additional shielding device at the top of the tank for loading the MOX fuels and additional safety plates to ensure safety are requested. The core is divided into two regions by inserting an inner tank to avoid criticality in MOX region only. The test region is composed by MOX fuel rods in the inner tank. Criticality is established by UO{sub 2} driver fuel rods outside of the inner tank. (Tsuchihashi, K.)

  11. Reactor physics needs in developing countries

    International Nuclear Information System (INIS)

    Solanilla, R.

    1980-01-01

    The aim of this paper the identification of needs on Reactor Physics in developing countries embarked in the installation and later on in the operation of Commercial Nuclear Power Plants. In this context the main task of Reactor Physics should be focused in the application of Physical models with inclusion of thermohydraulic process to solve the various realistic problems which appear to ensure a safe, economical and reliable core design and reactor operation. The first part of the paper deals with the scope of Reactor Physics and its interrelation with other disciplines as seen from the view point of developing countries possibilities. Needs requiring a quick response, i.e., those demands coming during the development of a specific Nuclear Power Plant Project, are summarized in the second part of the lecture. Plant startup has been chosen as reference to separate two categories of requirements: Requirements prior to startup phase include reactor core verification, licensing aspects review and study of fuel utilization alternatives; whereas the period during and after startup mainly embraces codes checkup and normalization, core follow-up and long term prediction

  12. Advanced methods in teaching reactor physics

    International Nuclear Information System (INIS)

    Snoj, Luka; Kromar, Marjan; Zerovnik, Gasper; Ravnik, Matjaz

    2011-01-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.

  13. Advanced methods in teaching reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Kromar, Marjan, E-mail: marjan.kromar@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Zerovnik, Gasper, E-mail: gasper.zerovnik@ijs.s [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Ravnik, Matjaz [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.

  14. Reactor physics calculations in the Nordic countries

    International Nuclear Information System (INIS)

    Hoeglund, R.

    1995-01-01

    The seventh biennial meeting on reactor physics calculations in the Nordic countries was arranged by VTT Energy on May 8-9, 1995. 26 papers on different subjects in the field of reactor physics were presented by 45 participants representing research establishments, technical universities, utilities, consultants and suppliers. Resent development and verification of the program systems of ABB Atom, Risoe, Scandpower, Studsvik and VTT Energy were the main topic of the meeting. Benchmarking of the two assembly codes CASMO-4 and HELIOS is proceeding. Cross section data calculated with CASMO-HEX have been validated for the Loviisa reactors. On core analysis ABB atom gives a description on its latest core simulator version POLCA7 with the calculation Core Master 2 and the BWR core supervision system Core Watch. Transient calculations with HEXTRAN, HEXTRAN- PLIM, TRAB, RAMONA, SIMULATE-3K and a code based on PRESTO II/POLCA7 were also presented

  15. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  16. Application of invariant embedding to reactor physics

    CERN Document Server

    Shimizu, Akinao; Parsegian, V L

    1972-01-01

    Application of Invariant Embedding to Reactor Physics describes the application of the method of invariant embedding to radiation shielding and to criticality calculations of atomic reactors. The authors intend to show how this method has been applied to realistic problems, together with the results of applications which will be useful to shielding design. The book is organized into two parts. Part A deals with the reflection and transmission of gamma rays by slabs. The chapters in this section cover topics such as the reflection and transmission problem of gamma rays; formulation of the probl

  17. A series of lectures on operational physics of power reactors

    International Nuclear Information System (INIS)

    Mohanakrishnan, P.; Rastogi, B.P.

    1982-01-01

    This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

  18. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    2004-03-01

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

  19. Reactor physics computations for nuclear engineering undergraduates

    International Nuclear Information System (INIS)

    Huria, H.C.

    1989-01-01

    The undergraduate program in nuclear engineering at the University of Cincinnati provides three-quarters of nuclear reactor theory that concentrate on physical principles, with calculations limited to those that can be conveniently completed on programmable calculators. An additional one-quarter course is designed to introduce the student to realistic core physics calculational methods, which necessarily requires a computer. Such calculations can be conveniently demonstrated and completed with the modern microcomputer. The one-quarter reactor computations course includes a one-group, one-dimensional diffusion code to introduce the concepts of inner and outer iterations, a cell spectrum code based on integral transport theory to generate cell-homogenized few-group cross sections, and a multigroup diffusion code to determine multiplication factors and power distributions in one-dimensional systems. Problem assignments include the determination of multiplication factors and flux distributions for typical pressurized water reactor (PWR) cores under various operating conditions, such as cold clean, hot clean, hot clean at full power, hot full power with xenon and samarium, and a boron concentration search. Moderator and Doppler coefficients can also be evaluated and examined

  20. NURESIM lecture on reactor physics (visual aids)

    International Nuclear Information System (INIS)

    Nguyen Tien Nguyen

    1998-01-01

    The purpose of the NURESIM software (NUclear REactor SIMulation) is to be used as a computer guide in quick view of the texts and pictures in the fields of nuclear reactor physics. This software is designed so that it can be used by users of different knowledge levels. Students could find here elementary concepts, researchers - important calculation codes as GRACE, PEACO, THERMOS, HEX120. The NURESIM software is composed of four parts: units, pictures, simulations and calculations. In the terminology of IAEA-TECDOC-314 (1984) the first three parts may be classified as a level 2 of sophistication IFM code package: ''Code package useful as a first introduction for nuclear engineers''. The last one (calculations) is classified as a level higher. Details about each part are explained in Paragraph 2. A users guide is in Paragraph 3. (author)

  1. Sterilization of swine wastewater treated by anaerobic reactors using UV photo-reactors

    Directory of Open Access Journals (Sweden)

    Erlon Lopes Pereira

    2014-09-01

    Full Text Available The use of ultraviolet radiation is an established procedure with growing application forthe disinfection of contaminated wastewater. This study aimed to evaluate the efficiency of artificial UV radiation, as a post treatment of liquid from anaerobic reactors treating swine effluent. The UV reactors were employed to sterilize pathogenic microorganisms. To this end, two photo-reactors were constructed using PVC pipe with100 mm diameter and 1060 mmlength, whose ends were sealed with PVC caps. The photo-reactors were designed to act on the liquid surface, as the lamp does not get into contact with the liquid. To increase the efficiency of UV radiation, photo-reactors were coated with aluminum foil. The lamp used in the reactors was germicidal fluorescent, with band wavelength of 230 nm, power of 30 Watts and manufactured by Techlux. In this research, the HRT with the highest removal efficiency was 0.063 days (90.6 minutes, even treating an effluent with veryhigh turbidity due to dissolved solids. It was concluded that the sterilization method using UV has proved to be an effective and appropriate process, among many other procedures.

  2. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    Keywords. Accelerator driven systems; nuclear waste transmutation; computer codes; reactor physics; reactor noise; kinetics; burnup; transport theory; Monte Carlo; thorium utilization; neutron multiplication; sub-criticality; sub-critical facilities.

  3. EBR-II Reactor Physics Benchmark Evaluation Report

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

    2017-12-28

    This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

  4. Common cause analysis of the TREAT upgrade reactor protection system

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.J.; Kamis, G.J.; Marbach, R.A.; Mueller, C.J.

    1984-09-01

    A triply redundant reactor scram system (RSS) has been designed for the upgraded TREAT facility. The independent failures reliability goal for the RSS is <10/sup -9/ failures per demand. An independent failures analysis indicated that this goal would be met. In addition, however, recognizing that in heavily redundant systems common-cause failures dominate, a common cause analysis of the TREAT upgrade RSS was done. The objective was to identify those common-cause initiators which could affect the functioning of the RSS, and to subsequently modify the design of the RSS so that the effect was minimized. A number of common-cause initiators were identified which were capable of defeating the triple redundancy feature of the reactor scram system. By means of a systematic analysis of the effect these initiators could have on the system, it was possible to identify seven necessary design and procedural modifications that would greatly reduce the probability of the reactor being run while the RSS was in a faulted condition.

  5. Multi-Physics Simulation of TREAT Kinetics using MAMMOTH

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, Mark; Gleicher, Frederick; Ortensi, Javier; Alberti, Anthony; Palmer, Todd

    2015-11-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific fuels transient tests range from simple temperature transients to full fuel melt accidents. The current TREAT core is driven by highly enriched uranium (HEU) dispersed in a graphite matrix (1:10000 U-235/C atom ratio). At the center of the core, fuel is removed allowing for the insertion of an experimental test vehicle. TREAT’s design provides experimental flexibility and inherent safety during neutron pulsing. This safety stems from the graphite in the driver fuel having a strong negative temperature coefficient of reactivity resulting from a thermal Maxwellian shift with increased leakage, as well as graphite acting as a temperature sink. Air cooling is available, but is generally used post-transient for heat removal. DOE and INL have expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility, with an emphasis on effective and safe operation while minimizing experimental time and cost. At INL, the Multi-physics Object Oriented Simulation Environment (MOOSE) has been selected as the model development framework for this work. This paper describes the results of preliminary simulations of a TREAT fuel element under transient conditions using the MOOSE-based MAMMOTH reactor physics tool.

  6. Biomass characteristics in three sequencing batch reactors treating a wastewater containing synthetic organic chemicals

    DEFF Research Database (Denmark)

    Hu, Z.Q.; Ferraina, R.A.; Ericson, J.F.

    2005-01-01

    The physical and biochemical characteristics of the biomass in three lab-scale sequencing batch reactors (SBR) treating a synthetic wastewater at a 20-day target solids retention time (SRT) were investigated. The synthetic wastewater feed contained biogenic compounds and 22 organic priming...... compounds, chosen to represent a wide variety of chemical structures with different N, P and S functional groups. At a two-day hydraulic retention time (HRT), the oxidation-reduction potential (ORP) cycled between -100 (anoxic) and 100mV (aerobic) in the anoxic/aerobic SBR, while it remained in a range...... of 126 +/- 18 and 249 +/- 18 mV in the aerobic sequencing batch biofilm reactor (SBBR) and the aerobic SBR reactor, respectively. A granular activated sludge with excellent settleability (SVI = 98 +/- 31 L mg(-1)) developed only in the anoxic/aerobic SBR, compared to a bulky sludge with poor settling...

  7. Advances in U.S. reactor physics standards

    International Nuclear Information System (INIS)

    Cokinos, Dimitrios

    2008-01-01

    The standards for Reactor Design, widely used in the nuclear industry, provide guidance and criteria for performing and validating a wide range of nuclear reactor calculations and measurements. Advances, over the past decades in reactor technology, nuclear data and infrastructure in the data handling field, led to major improvements in the development and application of reactor physics standards. A wide variety of reactor physics methods and techniques are being used by reactor physicists for the design and analysis of modern reactors. ANSI (American National Standards Institute) reactor physics standards, covering such areas as nuclear data, reactor design, startup testing, decay heat and fast neutron fluence in the pressure vessel, are summarized and discussed. These standards are regularly undergoing review to respond to an evolving nuclear technology and are being successfully used in the U.S and abroad contributing to improvements in reactor design, safe operation and quality assurance. An overview of the overall program of reactor physics standards is presented. New standards currently under development are also discussed. (authors)

  8. DOE fundamentals handbook: Nuclear physics and reactor theory

    International Nuclear Information System (INIS)

    1993-01-01

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance

  9. Performance of a UASB reactor treating coffee wet wastewater

    International Nuclear Information System (INIS)

    Guardia Puebla, Yans; Rodríguez Pérez, Suyén; Janet Jiménez Hernández; Sánchez Girón, Víctor

    2014-01-01

    The present work shows the results obtained in the anaerobic digestion process of coffee wet wastewater processing. An UASB anaerobic reactor was operated in single-stage in mesophilic temperature controlled conditions (37±1ºC). The effect of both organic loading rate (OLR) and hydraulic retention time (HRT) in the anaerobic digestion of coffee wet wastewater was investigated. The OLR values considered in the single-stage UASB reactor varied in a range of 3,6-4,1 kgCOD m-3 d-1 and the HRT stayed in a range of 21,5-15,5 hours. The evaluation results show that the best performance of UASB reactor in single-stage was obtained at OLR of 3,6 kg COD m-3 d-1 with an average value of total and soluble COD removal of 77,2% and 83,4%, respectively, and average methane concentration in biogas of 61%. The present study suggests that the anaerobic digestion is suitable to treating coffee wet wastewater. (author)

  10. Reactor physics in support of the naval nuclear propulsion programme

    International Nuclear Information System (INIS)

    Lisley, P.G.; Beeley, P.A.

    1994-01-01

    Reactor physics is a core component of all courses but in particular two postgraduate courses taught at the department in support of the naval nuclear propulsion programme. All of the courses include the following elements: lectures and problem solving exercises, laboratory work, experiments on the Jason zero power Argonaut reactor, demonstration of PWR behavior on a digital computer simulator and project work. This paper will highlight the emphasis on reactor physics in all elements of the education and training programme. (authors). 9 refs

  11. Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

    International Nuclear Information System (INIS)

    Wright, A.E.; Dutt, D.S.; Harrison, L.J.

    1990-01-01

    Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

  12. Communication and computer technologies for teaching physics in nuclear reactors

    International Nuclear Information System (INIS)

    Murua, C; Chautemps, A; Odetto, J; Keil, W; Trivino, S; Rossi, F; Perez Lucero, A

    2012-01-01

    In order to train personnel inn order to train personnel in Embalse Nuclear Power Plant, and provided that such training given primarily on the location of such a facility, we designed a pedagogical strategy that combined the use of conventional resources with new information technologies. Since the Nuclear Reactor RA-0 is an ideal tool for teaching Reactor Physics, priority was the use of it, both locally remotely. The teaching strategy is based on four pillar: -Lectures on the Power Plant (using a virtual classroom to support); -Remote monitoring of Ra-0 Nuclear Reactor parameters while operating (RA0REMOTO); -Use, through the Internet, of the Ra-0 Nuclear Reactor Simulator (RA0SIMUL); -Made in the Nuclear Reactor RA-0 of Reactor Physics practical. The work emphasizes RA0REMOTO and RA0SIMUL systems. The RA0REMOTO system is an appendix of the Electronic Data Acquisition System (SEAD) of the Nuclear Reactor RA-0. This system acquires signals from Reactor instrumentation and sends them to a server running the software that 'publish' the reactor parameters on the internet. Students may, during the lectures, monitor any parameter of the reactor while it operates, which allows teachers to compare theory with reality. RA0SIMUL is a simulator on the RA-0, which allows students to 'operate' a reactor analyzing the underlying physics concepts (author)

  13. Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems : Method Development for Multi-Physics in Reactors

    NARCIS (Netherlands)

    Perkó, Z.

    2015-01-01

    This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well

  14. Integral physics data for fast-reactor design

    International Nuclear Information System (INIS)

    Loewenstein, W.B.; Meneghetti, D.

    1962-01-01

    Integral physics data for fast-reactor design. The recent compilation of the section on fast-reactor physics for the forthcoming second edition of 'Reactor Physics Constants' has necessitated a survey of the available experimental integral data. The choice of fast-reactor-physics integral data to be included in the compilation was based upon two criteria besides availability: (a) the data arise from relatively simple systems which lend themselves to simple theoretical analyses; and (b) complicated systems representing prototypes or mock-ups having general interest in terms of fast-power reactors. The first criterion was decided upon so as to list integral data for those systems of most general utility for the verification of cross-section parameters and calculational procedures. The second criterion is based upon presentation of current data on actual fast power breeder reactor systems. These are too complicated for simple theoretical analysis. They demonstrate the complexity of the actual reactor versus the more idealized and easily analysed critical experiment. Integral physics data for reactor design refer to measurements on reactor systems, critical or otherwise, of the various reactor physics quantities of practical and/or theoretical importance. These characterize and lead to an understanding of the system. The measurements are represented by critical mass, core shape factor, detector ratios, neutron spectra, material replacement experiments, reflector savings, neutron lifetime, Rossi-α, and similar quantities. These data are reviewed and the range of applicability is described. Limitations of experimental and analytical results are shown to exist in certain spectral and criticality analyses. Experimental and analytical investigations are suggested for future work. These will tend to narrow the gap between theory and experiment on 'known' systems. They also include investigations to 'firm up' the physics of large conceptual, fast power-breeder reactor

  15. Annual progress report for 1982 of Theoretical Reactor Physics Section

    International Nuclear Information System (INIS)

    Rastogi, B.P.; Kumar, Vinod

    1983-01-01

    The progress of work done in the Theoretical Reactor Physics Section of the Bhabha Atomic Research Centre, Bombay, during the calendar year 1982 is reported in the form of write-ups and summaries. The main thrust of the work has been to master the neutronic design technology of four different types of nuclear reactor types, namely, pressurized heavy water reactors, boiling light water reactors, pressurized light water reactors and fast breeder reactors. The development work for the neutronic analysis, fuel design, and fuel management of the BWR type reactors of the Tarapur Atomic Power Station has been completed. A new reactor simulator system for PHWR design analysis and core follow-up was completed. Three dimensional static analysis codes based on nodal and finite element methods for the design work of larger size (500-750 MWe) reactors have been developed. Space link kinetics codes in one, two and three dimensions for above-mentioned reactor systems have been written and validated. Fast reactor core disruptive analysis codes have been developed. In the course of R and D work concerning various types of reactor projects, investigations were also carried in the allied areas of Monte Carlo techniques, integral transform methods, path integral methods, high spin states in heavy nuclei and a hydrodynamics model for a laser driven fusion system. (M.G.B.)

  16. Predictive modeling of coupled multi-physics systems: II. Illustrative application to reactor physics

    International Nuclear Information System (INIS)

    Cacuci, Dan Gabriel; Badea, Madalina Corina

    2014-01-01

    Highlights: • We applied the PMCMPS methodology to a paradigm neutron diffusion model. • We underscore the main steps in applying PMCMPS to treat very large coupled systems. • PMCMPS reduces the uncertainties in the optimally predicted responses and model parameters. • PMCMPS is for sequentially treating coupled systems that cannot be treated simultaneously. - Abstract: This work presents paradigm applications to reactor physics of the innovative mathematical methodology for “predictive modeling of coupled multi-physics systems (PMCMPS)” developed by Cacuci (2014). This methodology enables the assimilation of experimental and computational information and computes optimally predicted responses and model parameters with reduced predicted uncertainties, taking fully into account the coupling terms between the multi-physics systems, but using only the computational resources that would be needed to perform predictive modeling on each system separately. The paradigm examples presented in this work are based on a simple neutron diffusion model, chosen so as to enable closed-form solutions with clear physical interpretations. These paradigm examples also illustrate the computational efficiency of the PMCMPS, which enables the assimilation of additional experimental information, with a minimal increase in computational resources, to reduce the uncertainties in predicted responses and best-estimate values for uncertain model parameters, thus illustrating how very large systems can be treated without loss of information in a sequential rather than simultaneous manner

  17. Fast reactor physics at CEA: present studies and future prospects

    International Nuclear Information System (INIS)

    Hammer, P.

    1980-09-01

    This paper aims at giving a general survey of the fast reactor core physics and shielding studies wich are in progress at CEA (1979-1983) in order to solve the neutronic problems related to: - core design optimization, - reactor operation and fuel management, - safety, for the development of fast commercial breeders in France after the SUPER-PHENIX 1 construction is achieved

  18. Physical model study of neutron noise induced by vibration of reactor internals

    International Nuclear Information System (INIS)

    Liu Jinhui; Gu Fangyu

    1999-01-01

    The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

  19. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  20. On the research activities in reactor and neutron physics using the first egyptian research reactor

    International Nuclear Information System (INIS)

    Hassan, A.M.

    2000-01-01

    A review on the most important research activities in reactor and neutron physics using the first Egyptian Research Reactor (ET-RR-1) is given. An out look on: neutron cross-sections, neutron flux, neutron capture gamma-ray spectroscopy, neutron activation analysis, neutron diffraction and radiation shielding experiments, is presented

  1. Hamiltonian circuited simulations in reactor physics

    International Nuclear Information System (INIS)

    Rio Hirowati Shariffudin

    2002-01-01

    In the assessment of suitability of reactor designs and in the investigations into reactor safety, the steady state of a nuclear reactor has to be studied carefully. The analysis can be done through mockup designs but this approach costs a lot of money and consumes a lot of time. A less expensive approach is via simulations where the reactor and its neutron interactions are modelled mathematically. Finite difference discretization of the diffusion operator has been used to approximate the steady state multigroup neutron diffusion equations. The steps include the outer scheme which estimates the resulting right hand side of the matrix equation, the group scheme which calculates the upscatter problem and the inner scheme which solves for the flux for a particular group. The Hamiltonian circuited simulations for the inner iterations of the said neutron diffusion equation enable the effective use of parallel computing, especially where the solutions of multigroup neutron diffusion equations involving two or more space dimensions are required. (Author)

  2. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  3. The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G. [CEA/Saclay, DEN, 91 - Gif sur Yvette (France)] [and others

    2000-07-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  4. Multi-phased anaerobic baffled reactor treating food waste.

    Science.gov (United States)

    Ahamed, A; Chen, C-L; Rajagopal, R; Wu, D; Mao, Y; Ho, I J R; Lim, J W; Wang, J-Y

    2015-04-01

    This study was conducted to identify the performance of a multi-phased anaerobic baffled reactor (MP-ABR) with food waste (FW) as the substrate for biogas production and thereby to promote an efficient energy recovery and treatment method for the wastes with high organic solid content through phase separation. A four-chambered ABR was operated at an HRT of 30 days with an OLR of 0.5-1.0 g-VS/Ld for a period of 175 days at 35 ± 1°C. Consistent overall removal efficiencies of 85.3% (CODt), 94.5% (CODs), 89.6% (VFA) and 86.4% (VS) were observed throughout the experiment displaying a great potential to treat FW. Biogas generated was 215.57 mL/g-VS removed d. Phase separation was observed and supported by the COD and VFA trends, and an efficient recovery of bioenergy from FW was achieved. Copyright © 2015 Elsevier Ltd. All rights reserved.

  5. Physical Characteristics of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Ngo Quang Huy

    1994-10-01

    The operation of the TRIGA MARK II reactor of nominal power 250 KW has been stopped as all the fuel elements have been dismounted and taken away in 1968. The reconstruction of the reactor was accomplished with Russian technological assistance after 1975. The nominal power of the reconstructed reactor is of 500 KW. The recent Dalat reactor is unique of its kind in the world: Russian-designed core combined with left-over infrastructure of the American-made TRIGA II. The reactor was loaded in November 1983. It has reached physical criticality on 1/11/1983 (without central neutron trap) and on 18/12/1983 (with central neutron trap). The power start up occurred in February 1984 and from 20/3/1984 the reactor began to be operated at the nominal power 500 KW. The selected reports included in the proceedings reflect the start up procedures and numerous results obtained in the Dalat Nuclear Research Institute and the Centre of Nuclear Techniques on the determination of different physical characteristics of the reactor. These characteristics are of the first importance for the safe operation of the Dalat reactor

  6. Pebble Bed Reactor: core physics and fuel cycle analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Worley, B.A.

    1979-10-01

    The Pebble Bed Reactor is a gas-cooled, graphite-moderated high-temperature reactor that is continuously fueled with small spherical fuel elements. The projected performance was studied over a broad range of reactor applicability. Calculations were done for a burner on a throwaway cycle, a converter with recycle, a prebreeder and breeder. The thorium fuel cycle was considered using low, medium (denatured), and highly enriched uranium. The base calculations were carried out for electrical energy generation in a 1200 MW/sub e/ plant. A steady-state, continuous-fueling model was developed and one- and two-dimensional calculations were used to characterize performance. Treating a single point in time effects considerable savings in computer time as opposed to following a long reactor history, permitting evaluation of reactor performance over a broad range of design parameters and operating modes.

  7. Operating manual for the Health Physics Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1985-11-01

    This manual is intended to serve as a guide in the operation and maintenance of the Health Physics Researh Reactor (HPRR) of the Health Physics Dosimetry Applications Research (DOSAR) Facility. It includes descriptions of the HPRR and of associated equipment such as the reactor positioning devises and the derrick. Procedures for routine operation of the HPRR are given in detail, and checklists for the various steps are provided where applicable. Emergency procedures are similarly covered, and maintenance schedules are outlined. Also, a bibliography of references giving more detailed information on the DOSAR Facility is included. Changes to this manual will be approved by at least two of the following senior staff members: (1) the Operations Division Director, (2) the Reactor Operations Department Head, (3) the Supervisor of Reactor Operations TSF-HPRR Areas. The master copy and the copy of the manual issued to the HPRR Operations Supervisor will always reflect the latest revision. 22 figs.

  8. Computational mathematics and physics of fusion reactors.

    Science.gov (United States)

    Garabedian, Paul R

    2003-11-25

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors.

  9. Health physics in fusion reactor design

    International Nuclear Information System (INIS)

    Wong, K.Y.; Dinner, P.J.

    1984-06-01

    Experience in the control of tritium exposures to workers and the public gained through the design and operation of Ontario Hydro's nuclear stations has been applied to fusion projects and to design studies on emerging fusion reactor concepts. Ontario Hydro performance in occupational tritium exposure control and environmental impact is reviewed. Application of tritium control technologies and dose management methodology during facility design is highlighted

  10. Study of plutonium recycling physics in light water reactors

    International Nuclear Information System (INIS)

    Reuss, Paul

    1979-10-01

    A stock of plutonium from the reprocessing of thermal neutron reactor fuel is likely to appear in the next few years. The use of this plutonium as fuel replacing 235 U in thermal reactors is probably more interesting than simple stock-piling storage: immobilization of a capital which moreover would deteriorate by radioactive decay of isotope 241 also fissile and present to an appreciable extend in plutonium from reprocessing (half-life 15 years); recycling, on the other hand, will supply energy without complete degradation of the stock for fast neutron reactor loads, the burned matter having been partially renewed by conversion; furthermore the use of plutonium will meet the needs created by a temporary pressure on the naturel and/or enriched uranium market. For these two reasons the recycling of plutonium in thermal neutron reactors is being considered seriously today. The present work is confined to neutronic aspects and centres mainly on pressurized water-moderated reactors, the most highly developed at present in France. Four aspects of the problem are examined: 1. the physics of a plutonium-recycling reactor special features of neutronic phenomena with respect to the 'conventional' scheme of the 235 U burning reactor; 2. calculation of a plutonium-recycling reactor: adaptation of standard methods; 3. qualification of these calculations from the viewpoint of both data and inevitable approximations; 4. the fuel cycle and particularly the equivalence of fissile matters [fr

  11. Nuclear data and reactor physics activities in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Liem, P.H. [National Atomic Energy Agency, Tangerang (Indonesia). Center for Multipurpose Reactor

    1998-03-01

    The nuclear data and reactor physics activities in Indonesia, especially, in the National Atomic Energy Agency are presented. In the nuclear data field, the Agency is now taking the position of a user of the main nuclear data libraries such as JENDL and ENDF/B. These nuclear data libraries become the main sources for producing problem dependent cross section sets that are needed by cell calculation codes or transport codes for design, analysis and safety evaluation of research reactors. In the reactor physics field, besides utilising the existing core analysis codes obtained from bilateral and international co-operation, the Agency is putting much effort to self-develop Batan`s codes for reactor physics calculations, in particular, for research reactor and high temperature reactor design, analysis and fuel management. Under the collaboration with JAERI, Monte Carlo criticality calculations on the first criticality of RSG GAS (MPR-30) first core were done using JAERI continuous energy, vectorized Monte Carlo code, MVP, with JENDL-3.1 and JENDL-3.2 nuclear data libraries. The results were then compared with the experiment data collected during the commissioning phase. Monte Carlo calculations with both JENDL-3.1 and -3.2 libraries produced k{sub eff} values with excellent agreement with experiment data, however, systematically, JENDL-3.2 library showed slightly higher k{sub eff} values than JENDL-3.1 library. (author)

  12. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  13. Determination of the design excess reactivity for the TREAT Upgrade reactor

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Hanan, N.A.

    1983-01-01

    The excess reactivity designed to be built into a reactor core is a primary determinant of the fissile loadings of the fuel rods in the core. For the TREAT Upgrade (TU) reactor the considerations that enter into the determination of the excess reactivity are different from those of conventional power reactors. The reactor is designed to operate in an adiabatic transient mode for reactor safety in-pile test programs. The primary constituent of the excess reactivity is the calculated reactivity required to perform the most demanding transient experiments. Because of the unavailability of supporting critical experiments for the core design, the uncertainty terms that add on to this basic constituent are rather large. The burnup effects in TU are negligible and no refueling is planned. In this paper the determination of the design excess reactivity of the TREAT Upgrade reactor is discussed

  14. Neutrino Physics with Accelerator Driven Subcritical Reactors

    Science.gov (United States)

    Ciuffoli, Emilio

    2017-09-01

    Accelerator Driven Subcritical System (ADS) reactors are being developed around the world, to produce energy and, at the same time, to provide an efficient way to dispose of and to recycle nuclear waste. Used nuclear fuel, by itself, cannot sustain a chain reaction; however in ADS reactors the additional neutrons which are required will be supplied by a high-intensity accelerator. This accelerator will produce, as a by-product, a large quantity of {\\bar{ν }}μ via muon Decay At Rest (µDAR). Using liquid scintillators, it will be possible to to measure the CP-violating phase δCP and to look for experimental signs of the presence of sterile neutrinos in the appearance channel, testing the LSND and MiniBooNE anomalies. Even in the first stage of the project, when the beam energy will be lower, it will be possible to produce {\\bar{ν }}e via Isotope Decay At Rest (IsoDAR), which can be used to provide competitive bounds on sterile neutrinos in the disappearance channel. I will consider several experimental setups in which the antineutrinos are created using accelerators that will be constructed as part of the China-ADS program.

  15. The use of personal computers in reactor physics

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1988-01-01

    This paper points out that personal computers are now powerful enough (in terms of core size and speed) to allow them to be used for serious reactor physics applications. In addition the low cost of personal computers means that even small institutes can now have access to a significant amount of computer power. At the present time distribution centers, such as RSIC, are beginning to distribute reactor physics codes for use on personal computers; hopefully in the near future more and more of these codes will become available through distribution centers, such as RSIC

  16. Neutron physics of a high converting advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Berger, H.D.

    1985-01-01

    The neutron physics of an APWR are analysed by single pin-cell calculations as well as two-dimensional whole-reactor computations. The calculational methods of the two codes employed for this study, viz. the cell code SPEKTRA and the diffusion-burnup code DIBU, are presented in detail. The APWR-investigations carried out concentrate on the void coefficient characteristics of tight UO 2 /PuO 2 -lattices, control rod worths, burnup behaviour and spatial power distributions in APWR cores. The principal physics design differences between advanced pressurized water reactors and present-day PWRs are identified and discussed. (orig./HP) [de

  17. Reactor physics verification of the MCNP6 unstructured mesh capability

    International Nuclear Information System (INIS)

    Burke, T. P.; Kiedrowski, B. C.; Martz, R. L.; Martin, W. R.

    2013-01-01

    The Monte Carlo software package MCNP6 has the ability to transport particles on unstructured meshes generated from the Computed-Aided Engineering software Abaqus. Verification is performed using benchmarks with features relevant to reactor physics - Big Ten and the C5G7 computational benchmark. Various meshing strategies are tested and results are compared to reference solutions. Computational performance results are also given. The conclusions show MCNP6 is capable of producing accurate calculations for reactor physics geometries and the computational requirements for small lattice benchmarks are reasonable on modern computing platforms. (authors)

  18. KAMCCO, a reactor physics Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.

    1976-06-01

    KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de

  19. Stability and activity of anaerobic sludge from UASB reactors treating sewage in subtropical regions

    NARCIS (Netherlands)

    Seghezzo, L.; Cuevas, C.M.; Trupiano, A.P.; Guerra, R.G.; Gonzalez, S.M.; Zeeman, G.; Lettinga, G.

    2006-01-01

    The production of small amounts of well-stabilized biological sludge is one of the main advantages of upflow anaerobic sludge bed (UASB) reactors over aerobic wastewater treatment systems. In this work, sludge produced in three pilot-scale UASB reactors used to treat sewage under subtropical

  20. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, T.

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design

  1. International Conference on Physics and Technology of Reactors and Applications

    International Nuclear Information System (INIS)

    2007-01-01

    The first international conference on physics and technology of reactors and applications (PHYTRA 1) which took place in Marrakech (Morocco) from 14 to 16 March 2007, was designed to bring together scientists, teachers and students from universities, research centres and industry and other institutions to exchange knowledge and to discuss ideas and future issues. The programmes of the PHYTRA 1 conference covers a wide variety topics, the conference was organised in three plenary sessions, ten oral technical sessions and two poster sessions. The plenary sessions covers the following topics : The prospects of nuclear energy, The situation of nuclear sciences and energy in Morocco and Africa, and the new development in reactor physics and reactor design [fr

  2. Current Reactor Physics Benchmark Activities at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Margaret A. Marshall; Mackenzie L. Gorham; Joseph Christensen; James C. Turnbull; Kim Clark

    2011-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) [1] and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) [2] were established to preserve integral reactor physics and criticality experiment data for present and future research. These valuable assets provide the basis for recording, developing, and validating our integral nuclear data, and experimental and computational methods. These projects are managed through the Idaho National Laboratory (INL) and the Organisation for Economic Co-operation and Development Nuclear Energy Agency (OECD-NEA). Staff and students at the Department of Energy - Idaho (DOE-ID) and INL are engaged in the development of benchmarks to support ongoing research activities. These benchmarks include reactors or assemblies that support Next Generation Nuclear Plant (NGNP) research, space nuclear Fission Surface Power System (FSPS) design validation, and currently operational facilities in Southeastern Idaho.

  3. An Overview of the International Reactor Physics Experiment Evaluation Project

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gulliford, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-09

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

  4. Continuous fill intermittent decant type sequencing batch reactor application to upgrade the UASB treated sewage.

    Science.gov (United States)

    Khan, Abid Ali; Gaur, Rubia Zahid; Diamantis, Vasileios; Lew, Beni; Mehrotra, Indu; Kazmi, A A

    2013-05-01

    The performance of continuous flow intermittent decant type sequencing batch (CFID) reactor treating the effluent of an UASB reactor treating domestic wastewater and operated at 8 h hydraulic retention time (HRT) was investigated. The CFID was operated at three different HRTs (22, 8 and 6 h) and three different dissolved oxygen (DO) patterns (UASB systems, with a final effluent quality that comply with receiving water and effluent reuse criteria.

  5. Reactor physics research activities related to the very high temperature reactor in Japan

    International Nuclear Information System (INIS)

    Kaneko, Y.

    1987-01-01

    Reactor physics research activities in Japan that are related to the very high temperature reactor (VHTR) for multipurpose use are briefly summarized. Emphasis is placed on critical experiments. Neutronic core design accuracy required for the experimental VHTR is made clear, and nuclear data compilation and neutronic calculation code development are described. For experimental work, after a review of the results of all reactor physics experiments performed on the Semi-Homogeneous Experiment at the Japan Atomic Energy Research Institute, its reconstruction program to the VHTR critical assembly is presented. The aim of this program is to perform a detailed mockup experiment of the experimental VHTR loaded with low-enriched uranium-coated particle fuels. Finally, improvement of the neutronic calculation accuracy attained through comparison between calculation and experiment is illustrated, and some future problems are pointed out

  6. Methodology for reactor core physics analysis - part 2

    International Nuclear Information System (INIS)

    Ponzoni Filho, P.; Fernandes, V.B.; Lima Bezerra, J. de; Santos, T.I.C.

    1992-12-01

    The computer codes used for reactor core physics analysis are described. The modifications introduced in the public codes and the technical basis for the codes developed by the FURNAS utility are justified. An evaluation of the impact of these modifications on the parameter involved in qualifying the methodology is included. (F.E.). 5 ref, 7 figs, 5 tabs

  7. Successful vectorization - reactor physics Monte Carlo code

    International Nuclear Information System (INIS)

    Martin, W.R.

    1989-01-01

    Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)

  8. Anaerobic biogranulation in a hybrid reactor treating phenolic waste

    International Nuclear Information System (INIS)

    Ramakrishnan, Anushyaa; Gupta, S.K.

    2006-01-01

    Granulation was examined in four similar anaerobic hybrid reactors 15.5 L volume (with an effective volume of 13.5 L) during the treatment of synthetic coal wastewater at the mesophilic temperature of 27 ± 5 deg. C. The hybrid reactors are a combination of UASB unit at the lower part and an anaerobic filter at the upper end. Synthetic wastewater with an average chemical oxygen demand (COD) of 2240 mg/L, phenolics concentration of 752 mg/L and a mixture of volatile fatty acids was fed to three hybrid reactors. The fourth reactor, control system, was fed with a wastewater containing sodium acetate and mineral nutrients. Coal waste water contained phenol (490 mg/L); m-, o-, p-cresols (123.0, 58.6, 42 mg/L); 2,4-, 2,5-, 3,4- and 3,5-dimethyl phenols (6.3, 6.3, 4.4 and 21.3 mg/L) as major phenolic compounds. A mixture of anaerobic digester sludge and partially granulated sludge (3:1) were used as seed materials for the start up of the reactors. Granules were observed after 45 days of operation of the systems. The granules ranged from 0.4 to 1.2 mm in diameter with good settling characteristics with an SVI of 12 mL/g SS. After granulation, the hybrid reactor performed steadily with phenolics and COD removal efficiencies of 93% and 88%, respectively at volumetric loading rate of 2.24 g COD/L d and hydraulic retention time of 24 h. The removal efficiencies for phenol and m/p-cresols reached 92% and 93% (corresponding to 450.8 and 153 mg/L), while o-cresol was degraded to 88% (corresponding to 51.04 mg/L). Dimethyl phenols could be removed completely at all the organic loadings and did not contribute much to the residual organics. Biodegradation of o-cresol was obtained in the hybrid-UASB reactors

  9. Physics and engineering aspects of the EBT reactor

    International Nuclear Information System (INIS)

    Uckan, N.A.; Bettis, E.S.; Hedrick, C.L.; Santoro, R.T.; Watts, H.L.; Yeh, H.T.

    1977-01-01

    The ELMO Bumpy Torus (EBT) reactor has the advantage of high-β, steady-state operation. The first reactor study based on the EBT confinement concept was initiated in 1976. It provided the required starting point for continued assessment of the validity of the concept. A new design based on the present physics understanding, practical design approaches, and present and near-term technologies has been established. One of the important factors in an EBT reactor is the large aspect ratio (large toroidal major radius as well). This leads to a power plant with a comparatively large total energy output, usually in the range of 2000-6000 MW(th) for a conventional neutron wall loading of 1-2 MW/m 2 (the high value of β in an EBT device provides a net cost per unit energy roughly equal to or somewhat less than that for a Tokamak system). The large aspect ratio also provides very simple engineering and design requirements because of good access and small force loading asymmetries. Another important factor is the steady-state operation. In an EBT system, less power handling, energy storage, and filtering equipment will be needed. An EBT reactor is less likely to be subject to thermal and mechanical fatigue than reactors with large pulsed magnetic fields and short bursts of fusion power. The details of the key design elements and critical scientific and technology factors which are substantially different from other fusion reactor approaches are described

  10. Nuclear data and integral experiments in reactor physics

    International Nuclear Information System (INIS)

    Farinelli, U.

    1980-01-01

    The material given here broadly covers the content of the 10 lectures delivered at the Winter Course on Reactor Theory and Power Reactors, ICTP, Trieste (13 February - 10 March 1978). However, the parts that could easily be found in the current literature have been omitted and replaced with the appropriate references. The needs for reactor physics calculations, particularly as applicable to commercial reactors, are reviewed in the introduction. The relative merits and shortcomings of fundamental and semi-empirical methods are discussed. The relative importance of different nuclear data, the ways in which they can be measured or calculated, and the sources of information on measured and evaluated data are briefly reviewed. The various approaches to the condensation of nuclear data to multigroup cross sections are described. After some consideration to the sensitivity calculations and the evaluation of errors, some of the most important type of integral experiments in reactor physics are introduced, with a view to showing the main difficulties in the interpretation and utilization of their results and the most recent trends in experimentation. The conclusions try to assign some priorities in the implementation of experimental and calculational capabilities, especially for a developing country. (author)

  11. Proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic

    International Nuclear Information System (INIS)

    1986-01-01

    The proceedings of the 6. National Meeting of Reactor Physics and Thermohydraulic - 6. ENFIR - allow to evaluate the present status of development in reactor physics and thermohydraulic fields. The mathematical models and methods for calculating neutronic of nuclear reactors, safety reactor analysis, measuring methods of neutronic parameters, computerized simulation of accidents, transients and thermohydraulic analysis are presented. (M.C.K.) [pt

  12. Reactor physics innovations of the advanced CANDU reactor core: adaptable and efficient

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Hopwood, J.M.; Bonechi, M.

    2003-01-01

    The Advanced CANDU Reactor (ACR) is designed to have a benign, operator-friendly core physics characteristic, including a slightly negative coolant-void reactivity and a moderately negative power coefficient. The discharge fuel burnup is about three times that of natural uranium fuel in current CANDU reactors. Key features of the reactor physics innovations in the ACR core include the use of H 2 O coolant, slightly enriched uranium (SEU) fuel, and D 2 O moderator in a reduced lattice pitch. These innovations result in substantial improvements in economics, as well as significant enhancements in reactor performance and waste reduction over the current reactor design. The ACR can be readily adapted to different power outputs by increasing or decreasing the number of fuel channels, while maintaining identical fuel and fuel-channel characteristics. The flexibility provided by on-power refuelling and simple fuel bundle design enables the ACR to easily adapt to the use of plutonium and thorium fuel cycles. No major modifications to the basic ACR design are required because the benign neutronic characteristics of the SEU fuel cycle are also inherent in these advanced fuel cycles. (author)

  13. Physical modalities for treating acne and rosacea.

    Science.gov (United States)

    Jalian, H Ray; Levin, Yakir; Wanner, Molly

    2016-06-01

    Physical modalities provide an important adjunct to medical treatment of acne and rosacea. In patients who cannot tolerate or fail medical treatments, physical modalities offer an alternative approach. For cases of acne scarring, phymatous changes of rosacea, and rosacea-associated telangiectasia, physical modalities such as laser and light treatments represent the treatment of choice. We will review the use of laser and light treatments, photodynamic therapy, and other physical modalities such as targeted therapies for the treatment of acne and rosacea. ©2016 Frontline Medical Communications.

  14. Neutron physics for nuclear reactors unpublished writings by Enrico Fermi

    CERN Document Server

    Fermi, Enrico; Pisanti, O

    2010-01-01

    This unique volume gives an accurate and very detailed description of the functioning and operation of basic nuclear reactors, as emerging from yet unpublished papers by Nobel Laureate Enrico Fermi. In the first part, the entire course of lectures on Neutron Physics delivered by Fermi at Los Alamos is reported, according to the version made by Anthony P French. Here, the fundamental physical phenomena are described very clearly and comprehensively, giving the appropriate physics grounds for the functioning of nuclear piles. In the second part, all the patents issued by Fermi (and coworkers) on

  15. Proceedings on the Second Autumn School on Reactor Physics EROEFI II

    International Nuclear Information System (INIS)

    Racz, A.

    1995-01-01

    The main topics of the Reactor Physics School were neutron and reactor physical calculations, reactor safety, systems theory, simulation of accidents, reactor monitoring system, computer codes and procedures for solving specific problems in the field of nuclear reactors (especially safety). A special attention was paid to the AGNES project. Papers falling in the INIS scope have been abstracted and indexed individually for the INIS database. (K.A.)

  16. Photocatalytic reactors for treating water pollution with solar illumination. II: a simplified analysis for flow reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Inst. fuer Technische Chemie, Univ. Hannover, Hannover (Germany); Brandi, R.J.; Cassano, A.E. [INTEC (Univ. Nacional del Litoral and CONICET), Santa Fe (Argentina)

    2003-07-01

    Very frequently outgoing streams of real wastewaters do not have a definite and constant composition. Additionally, when the degradation process makes use of solar irradiation, the photon flux is hardly constant. These two factors strongly militate against the use of very elaborate, exact models for analyzing the performance of the employed reactors. In these cases, approximate methods may be the most practical approach. One possible way is presented in this paper. The observed photonic efficiency concept developed in a previous contribution (sagawe et al., 2002a) is applied to continuous reactors for both steady state and transient operations of photocatalytic reactions applied to wastewaters decontamination processes. For this reactor the local observed photonic efficiency, defined at each reactor longitudinal position, is the convenient property to express the concentration spatial evolution. It is also shown that the description of the reactor performance employing a mass balance can be done in a rather simple way introducing a mass-moving coordinate transformation that remodel the mass inventory and permits working with simpler ordinary differential equations. (orig.)

  17. Health physics aspects of a research reactor fuel shipment

    International Nuclear Information System (INIS)

    Dodd, B.; Johnson, A.G.; Anderson, T.V.

    1984-01-01

    In June 1982, 92 irradiated fuel elements were shipped from the Oregon State University TRIGA Reactor to Westinghouse Hanford Corporation to be used in the Fuel Materials Examination Facility, This paper describes some of the health physics aspects of the planning, preparation and procedures associated with that shipment. In particular, the lessons learned are described in order that the benefits of the experience gained may be readily available to other small institutions. (author)

  18. Production test-080, physics testing at D reactor deactivation

    Energy Technology Data Exchange (ETDEWEB)

    Bailey, G.F.

    1967-06-15

    The purpose of this test is to provide a set of experimental data to test a compute code frequently used in nuclear safety analyses and to explore certain experimental techniques which may prove extremely valuable in the future. In addition, some basic physics parameters which will be measured may be used in an assessment of the feasibility of using a deactivated Hanford reactor for space-dependent transient tests.

  19. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  20. Summary of ORSphere Critical and Reactor Physics Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Margaret A.; Bess, John D.

    2016-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.

  1. Summary of ORSphere critical and reactor physics measurements

    Directory of Open Access Journals (Sweden)

    Marshall Margaret A.

    2017-01-01

    Full Text Available In the early 1970s Dr. John T. Mihalczo (team leader, J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF with highly enriched uranium (HEU metal (called Oak Ridge Alloy or ORALLOY to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP. Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

  2. Summary of ORSphere critical and reactor physics measurements

    Science.gov (United States)

    Marshall, Margaret A.; Bess, John D.

    2017-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

  3. Study of Physical Protection System at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ligam, A.S.; Ina, I.; Zarina Masood

    2016-01-01

    Physical protection program at PUSPATI TRIGA Reactor (RTP) which is located at Nuklear Malaysia, Bangi Complex has been strengthened and upgraded from time to time to accommodate current situation needs. However, there is always room for improvement. Hence, study have been made to look deeper into physical protection components such as delay systems, external sensors, PPS intruder alarm sensors, use of video system, personnel security or insider threats, access control operation system operation rules and security culture that may need to take into consideration. (author)

  4. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  5. Physics and Material Problems of Reactor Control Rods. Proceedings of the Symposium on Physics and Material Problems of Reactor Control Rods

    International Nuclear Information System (INIS)

    1964-01-01

    The development of nuclear reactors is closely associated with the progress made in the solution of control problems. To survey the present state of the subject the International Atomic Energy Agency convened a symposium devoted to ''Physics and Material Problems in Reactor Control Rods''. The Symposium was held in Vienna from 11 to 15 November 1963 and was attended by more than 100 participants representing 21 of the Agency's Member States and two international organizations. Problems discussed in the 34 papers presented at 8 sessions covered many special aspects of theoretical and experimental physics, engineering, metallurgy, etc. The first session of the Symposium was devoted to different theoretical methods used for the determination of control rod effectiveness in a multi- regioned reactor, and in natural-uranium heavy-water moderated cores. Homogeneous and heterogeneous approaches were discussed and applicability of proposed methods for various forms of control elements considered. During the two following sessions a number of theoretical problems and mathematical models were examined together with various control rod experiments and measurements in exponential and critical assemblies and at commercial nuclear power stations. The next session dealt with the connection between physics and technology of control rods, the latter being the subject of the remainder of the Symposium. Testing and actual operating experience of control rods were also treated in some of the presented papers. The session on engineering aspects of control rod systems included presentation of research results in a marine control station, the design of large graphite reactor control drives and the description of different mechanisms for rapid insertion of control absorbers. Finally, the methods of fast reactor control were discussed, followed by the presentation of various ''unconventional'' methods of reactivity control, such as hydraulic ball, fluidized bed, gas pressure and soluble

  6. IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

    International Nuclear Information System (INIS)

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

  7. Multi-physic simulations of irradiation experiments in a technological irradiation reactor

    International Nuclear Information System (INIS)

    Bonaccorsi, Th.

    2007-09-01

    A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

  8. Nuclear energy renaissance and reactor physics. Enlightenment of PHYSOR'08

    International Nuclear Information System (INIS)

    Peng Feng

    2010-01-01

    In relation to world's growing energy demands and concerns on global warming, nuclear energy as a sustainable resource is in its new period of renaissance. This is reflected in the record number of 447 papers on the International Conference on the Physics of Reactors--PHYSOR'08 held in Switzerland in 2008. The contents of these papers include the developments and frontiers in various directions of reactor physics. Featured by vast area of subjects, these emphasize the fact that the scope of the reactor physicist's R and D interests has expands considerably in recent years. The main keynote addresses and technical plenary lectures are briefly introduced. Some items concerned by the conference, such as: the status and perspective of nuclear energy's R and D, deployment and policy in main nuclear nations, the potential role of nuclear energy in mitigation global warming and slow down the GHG release, the sustainability of resource for nuclear energy utilization. Status and outlook about the needs of research and test facilities required in nuclear energy development, etc. are discussed. (authors)

  9. The reactor physics computer programs in PC's era

    International Nuclear Information System (INIS)

    Nainer, O.; Serghiuta, D.

    1995-01-01

    The main objective of reactor physics analysis is the evaluation of flux and power distribution over the reactor core. For CANDU reactors sophisticated computer programs, such as FMDP and RFSP, were developed 20 years ago for mainframe computers. These programs were adapted to work on workstations with UNIX or DOS, but they lack a feature that could improve their use and that is 'user friendly'. For using these programs the users need to deal with a great amount of information contained in sophisticated files. To modify a model is a great challenge. First of all, it is necessary to bear in mind all the geometrical dimensions and accordingly, to modify the core model to match the new requirements. All this must be done in a line input file. For a DOS platform, using an average performance PC system, could it be possible: to represent and modify all the geometrical and physical parameters in a meaningful way, on screen, using an intuitive graphic user interface; to reduce the real time elapsed in order to perform complex fuel-management analysis 'at home'; to avoid the rewrite of the mainframe version of the program? The author's answer is a fuel-management computer package operating on PC, 3 time faster than on a CDC-Cyber 830 mainframe one (486DX/33MHz/8MbRAM) or 20 time faster (Pentium-PC), respectively. (author). 5 refs., 1 tab., 5 figs

  10. Physics with Reactor Neutrinos: The Show has Begun!

    Science.gov (United States)

    Winslow, Lindley

    2012-03-01

    The next generation of reactor neutrino experiments started with the first Double Chooz result earlier this year and will continue with the RENO and Daya Bay experiments. The main goal of these experiments is the search for the last unknown mixing angle governing neutrino oscillations θ13. The results of these experiments will complete our picture of neutrino oscillations and are key for planning searches for CP violation. Along the way, they may teach us something about sterile neutrinos and the application of neutrinos to issues of nuclear non-proliferation. The physics and design of reactor neutrino experiments will be discussed, especially as it relates to the Double Chooz measurement of θ13.

  11. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy

    2000-04-01

    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  12. Research on reactor physics using the Very High Temperature Reactor Critical Assembly (VHTRC)

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1988-01-01

    The High Temperature Engineering Test Reactor (HTTR), of which the research and development are advanced by Japan Atomic Energy Research Institute, is planned to apply for the permission of installation in fiscal year 1988, and to start the construction in the latter half of fisical year 1989. As the duty of reactor physics research, the accuracy of the nuclear data is to be confirmed, the validity of the nuclear design techniques is to be inspected, and the nuclear safety of the HTTR core design is to be verified. Therefore, by using the VHTRC, the experimental data of the reactor physics quantities are acquired, such as critical mass, the reactivity worth of simulated control rods and burnable poison rods, the temperature factor of reactivity, power distribution and so on, and the experiment and analysis are advanced. The cores built up in the VHTRC so far were three kinds having different lattice forms and degrees of uranium enrichment. The calculated critical mass was smaller by 1-5 % than the measured values. As to the power distribution and the reactivity worth of burnable poison rods, the prospect of satisfying the required accuracy for the design of the HTTR core was obtained. The experiment using a new core having axially different enrichment degree is planned. (K.I.)

  13. International Reactor Physics Experiment Evaluation (IRPhE) Project

    International Nuclear Information System (INIS)

    2013-01-01

    The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

  14. Modelling of sludge blanket height and flow pattern in UASB reactors treating municipal wastewater

    International Nuclear Information System (INIS)

    Singh, K.S.; Viraraghavan, T.

    2002-01-01

    Two upflow anaerobic sludge blanket (UASB) reactors were started-up and operated for approximately 900 days to examine the feasibility of treating municipal wastewater under low temperature conditions. A modified solid distribution model was formulated by incorporating the variation of biogas production rate with a change in temperature. This model was used to optimize the sludge blanket height of UASB reactors for an effective operation of gas-liquid-solid (GLS) separation device. This model was found to simulate well the solid distribution as confirmed experimental observation of solid profile along the height of the reactor. Mathematical analysis of tracer curves indicated the presence of a mixed type of flow pattern in the sludge-bed zone of the reactor. It was found that the dead-zone and by-pass flow fraction were impacted by the change in operating temperatures. (author)

  15. CFD Simulation of an Anaerobic Membrane BioReactor (AnMBR to Treat Industrial Wastewater

    Directory of Open Access Journals (Sweden)

    Laura C. Zuluaga

    2015-06-01

    Full Text Available A Computational Fluid Dynamics (CFD simulation has been developed for an Anaerobic Membrane BioReactor (AnMBR to treat industrial wastewater. As the process consists of a side-stream MBR, two separate simulations were created: (i reactor and (ii membrane. Different cases were conducted for each one, so the surrounding temperature and the total suspended solids (TSS concentration were checked. For the reactor, the most important aspects to consider were the dead zones and the mixing, whereas for the ceramic membrane, it was the shear stress over the membrane surface. Results show that the reactor's mixing process was adequate and that the membrane presented higher shear stress in the 'triangular' channel.

  16. Proceedings of the 10. Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    Santos Bastos, W. dos

    1995-01-01

    These proceedings presents all the Meeting papers emphasizing specific aspects on reactor physics method, criticality, fuel management, nuclear data, safety analysis, simulation and shielding, neutronics, thermal hydraulics, reactor operation and computational methods

  17. Application of an innovative methodology to improve the starting-up of UASB reactors treating domestic sewage.

    Science.gov (United States)

    Rodríguez, J A; Peña, M R; Manzi, V

    2001-01-01

    This study shows the results obtained during the starting-up evaluation of an UASB reactor treating domestic sewage. It is located in the municipality of Ginebra, Valle del Cauca region in Colombia. Its design flow is 7.5 l/s with a maximum capacity of 10 l/s. The reactor was seeded with a deficient quality inoculum which accounted for 20% of the total reactor volume. The starting-up methodology comprised the sequential washing of the sludge (inoculum) by applying three different upflow velocities. This procedure resembles what other authors term the "selective pressure method". Once the sludge was washed, the reactor was started-up with an initial hydraulic retention time (HRT) of 24.9 hours, which was steadily reduced down to 6.7 hours in the final stage. Along the starting-up phase, there was a positive evolution in terms of quantity, quality and spatial distribution of the sludge. Consequently, there was a positive evolution in organic matter removal mechanisms. For HRT above 14 hours, the removal mechanisms were mainly physical whilst for HRT below 9 hours the removal mechanisms were mostly biological. Based on the above considerations and on the water quality parameters measured, it may be concluded that the start-up of an UASB reactor for domestic sewage treatment seeded with a low quality inoculum can be done with HRT as low as 15 or 12 hours. In this way, it is possible to reduce the starting-up period of these reactors down to 4 to 6 weeks, provided that the starting-up methodology is properly applied.

  18. Proceedings of the 1992 topical meeting on advances in reactor physics

    International Nuclear Information System (INIS)

    1992-01-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements ampersand Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  19. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  20. Using Vega Linux Cluster at Reactor Physics Dept

    International Nuclear Information System (INIS)

    Zefran, B.; Jeraj, R.; Skvarc, J.; Glumac, B.

    1999-01-01

    Experience using a Linux-based cluster for the reactor physics calculations are presented in this paper. Special attention is paid to the MCNP code in this environment and to practical guidelines how to prepare and use the paralel version of the code. Our results of a time comparison study are presented for two sets of inputs. The results are promising and speedup factor achieved on the Linux cluster agrees with previous tests on other parallel systems. We also tested tools for parallelization of other programs used at our Dept..(author)

  1. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  2. Reactor Physics Characterization of the HTR Module with UCO Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gerhard Strydom

    2011-01-01

    ABSTRACT The HTR Module [1] is a graphite-moderated, helium cooled pebble bed High Temperature Reactor (HTR) design that has been extensively used as a reference template for the former South African and current Chinese HTR [2] programs. This design utilized spherical fuel elements packed into a dynamic pebble bed, consisting of TRISO coated uranium oxide (UO2) fuel kernels with a U-235 enrichment of 7.8% and a Heavy Metal loading of 7 grams per pebble. The main objective of this study is to compare several important reactor physics and core design parameters for the HTR Module and an identical design utilizing UCO fuel kernels. Fuel kernels of this type are currently being tested in the Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR) as part of the larger Next Generation Nuclear Plant (NGNP) project. The PEBBED-THERMIX [3] code, which was developed specifically for the analysis of pebble bed HTRs, was used to compare the coupled neutronic and thermal fluid performance of the two designs.

  3. Proceedings of the nineteenth symposium of atomic energy research on WWER reactor physics and reactor safety

    International Nuclear Information System (INIS)

    Vidovszky, I.

    2009-10-01

    The present volume contains 55 papers, presented on the nineteenth symposium of atomic energy research, held in Varna, Bulgaria, 21-25 September 2009. The papers are presented in their original form, i. e. no corrections or modifications were carried out. The content of this volume is divided into thematic groups: Fuel Management, Spectral and Core Calculations, Core Surveillance and Monitoring, CFD Analysis, Reactor Dynamics Thermal Hydraulics and Safety Analysis, Physical Problems of Spent Fuel Decommissioning and Radwaste, Actinide Transmutation and Spent Fuel Disposal, Core Operation, Experiments and Code Validation - according to the presentation sequence on the Symposium. (Author)

  4. Progress report on reactor physics research program, January 1963 - February 1964

    International Nuclear Information System (INIS)

    1964-02-01

    This progress report is a part of the annual report of the department of reactor physics prepared for the Boris Kidric Institute of nuclear sciences. It is a review of research activities in the field of theoretical and experimental reactor physics in the year 1973. A part of this program was included in the NPY Cooperative program in reactor physics. The topics covered by this report are as follows: Calculations of the thermal neutron distribution and reaction rate in a reactor cell and comparison with experiments; buckling measurements; thermalization and slowing down of neutrons; pulsed neutron source techniques; and reactor kinetics

  5. Mathematical modeling of upflow anaerobic sludge blanket (UASB) reactor treating domestic wastewater.

    Science.gov (United States)

    Elmitwalli, Tarek

    2013-01-01

    Although the upflow anaerobic sludge blanket (UASB) reactor has been widely applied for domestic wastewater treatment in many developing countries, there is no sufficient mathematical model for proper design and operation of the reactor. An empirical model based on non-linear regression was developed to represent the physical and chemical removal of suspended solids (SS) in the reactor. Moreover, a simplified dynamic model based on ADM1 and the empirical model for SS removal was developed for anaerobic digestion of the entrapped SS and dissolved matter in the wastewater. The empirical model showed that effluent suspended chemical oxygen demand (COD(ss)) concentration is directly proportional to the influent COD(ss) concentration and inversely proportional to both the hydraulic retention time (HRT) of the reactor and wastewater temperature. For obtaining sufficient COD(ss) removal, the HRT of the UASB reactor must be higher than 4 h, and higher HRT than 12 h slightly improved COD(ss) removal. The dynamic model results showed that the required time for filling the reactor with sludge mainly depends on influent total chemical oxygen demand (COD(t)) concentration and HRT. The influent COD(t) concentration, HRT and temperature play a crucial role on the performance of the reactor. The results indicated that shorter HRT is needed for optimization of COD(t) removal, as compared with optimization of COD(t) conversion to methane. Based on the model results, the design HRT of the UASB reactor should be selected based on the optimization of wastewater conversion and minimization of biodegradable SS accumulation in the sludge bed, not only based on COD removal, to guarantee a stable reactor performance.

  6. Initial Testing of the Microscopic Depletion Implementation in the MAMMOTH Reactor Physics Application

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ganapol, B. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, F. N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, B. [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, M. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Present and new nuclear fuels that will be tested at the Transient Reactor Test (TREAT) facility will be analyzed with the MAMMOTH reactor physics application, currently under development, at Idaho National Laboratory. MAMMOTH natively couples the BISON, RELAP-7, and Rattlesnake applications within the MOOSE framework. This system allows the irradiation of fuel from beginning of life in a nuclear reactor until it is placed in TREAT for fuel testing within the same analysis mesh and, thus, retaining a very high level of resolution and fidelity. The calculation of the isotopic distribution in fuel requires the solution to the decay and transmutation equations coupled to the neutron transport equation. The Chebyshev Rational Approximation Method (CRAM) is the current state-of-the-art in the field, as was chosen to be the solver for the decay and transmutation equations. This report shows that the implementation of the CRAM solver within MAMMOTH is correct with various analytic benchmarks for decay and transmutation of nuclides. The results indicate that the solutions with CRAM order 16 achieve the level of precision of the benchmark. The CRAM solutions show little sensitivity to the time step size and consistently produce a high level of accuracy for isotopic decay for time steps of 1x10^11 years. Comparisons to DRAGON5 with 297 isotopes yield comparable results, but some differences need to be further analyzed.

  7. Methodology and results of investigations of physical parameters of high-temperature reactors

    International Nuclear Information System (INIS)

    Cherepnin, Yu.S.; Chertkov, Yu.B.

    1995-01-01

    A physical investigations of reactors of stand complexes Baikal-1 and IGR have been carrying out more 30 years. Measuring methods of the physical investigations were divided into 2 groups: 1) methods for measuring of reactivity effects; 2) methods for measuring relative and absolute values of neutron flux and power release. The physical investigations on the reactors IVG-1 and IGR were carryied out under following conditions: during physical starts-up of regular variants of reactor cores; during energy starts-up of the reactors; before beginning of new loop chanel tests of the reactors; during research hot starts-up of the reactors the physical parameters were controled. The most full and authentic information about studied reactor have been providing by physical investigations. In 1984 physical investigations were carryied out on the IGR reactor and then the hot start-up of the mostest power and mostest large on fuel loading loop chanel was carryied out. This chanel contained 6 fuel assemblies with the summary fuel loading 3,06 kilogrammes of uranium and it was calculated for power equal to 20 MW. In 1988 the physical investigations for selection of project process chanels destined for new water cooled reactor core were carryied out. In 1993 the neutron-physical calculation on possibility of tests for the rector Nerva fuel element was carryied out. 9 refs., 4 figs

  8. Conceptual study of fusion-driven transmutation reactor with ITER physics and engineering constraints

    Science.gov (United States)

    Hong, Bong

    2011-10-01

    A conceptual study of fusion-driven transmutation reactor was performed based on ITER physics and engineering constraints. A compact reactor concept is desirable from an economic viewpoint. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which inter-relate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor; the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burnup calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor.

  9. Considerations in treating physically active older adults and aging athletes.

    Science.gov (United States)

    Langer, Paul R

    2015-04-01

    Life spans are increasing and research is showing more and more how important exercise is to successful aging. Medical practitioners need to appreciate the physiologic and physical changes that occur with age, as well as the significant benefits of physical activity, so they not only can properly treat their older patients but also so they can promote the benefits of exercise to their sedentary older patients. Copyright © 2015 Elsevier Inc. All rights reserved.

  10. Physical Activity in Patients Treated With Peritoneal Dialysis

    Directory of Open Access Journals (Sweden)

    Tharshika Thangarasa

    2017-03-01

    Full Text Available Background: Patients with chronic diseases are known to benefit from exercise. Despite a lack of compelling evidence, patients with end-stage kidney disease treated with peritoneal dialysis are often discouraged from participating in exercise programs that include resistance training due to concerns about the development of hernias and leaks. The actual effects of physical activity with or without structured exercise programs for these patients remain unclear. The purpose of this study is to more completely define the risks and benefits of physical activity in the end-stage kidney disease population treated with peritoneal dialysis. Methods/design: We will conduct a systematic review examining the effects of physical activity on end-stage kidney disease patients treated with peritoneal dialysis. For the purposes of this review, exercise will be considered a purposive subcategory of physical activity. The primary objective is to determine if physical activity in this patient population is associated with improvements in mental health, physical functioning, fatigue and quality of life and if there is an increase in adverse outcomes. With the help of a skilled librarian, we will search MEDLINE, EMBASE, CINAHL, and Cochrane Central Register of Controlled Trials for randomized trials and observational studies. We will include adult end-stage kidney disease patients treated with peritoneal dialysis that have participated in an exercise training program or had their level of physical activity assessed directly or by self-report. The study must include an assessment of the association between physical activity and one of our primary or secondary outcomes measures. We will report study quality using the Cochrane Risk of Bias Assessment Tool for randomized controlled trials and the Newcastle–Ottawa Scale for observational studies. Quality across studies will be assessed using the Grading of Recommendations Assessment, Development and Evaluation (GRADE

  11. Microorganism selection and performance in bioslurry reactors treating PAH-contaminated soil.

    Science.gov (United States)

    Cassidy, D P; Hudak, A J

    2002-09-01

    A continuous-flow reactor (CSTR) and a soil slurry-sequencing batch reactor (SS-SBR) were operated in 81 vessels for 200 days to treat a soil contaminated with polycyclic aromatic hydrocarbons (PAH). Filtered slurry samples were used to quantify bulk biosurfactant concentrations and PAH emulsification. Concentrations of Corynebacterium aquaticum, Flavobacterium mizutaii, Mycobacterium gastri, Pseudomonas aeruginosa, and Pseudomonas putida were determined using fatty acid methyl ester (FAME) analysis. The CSTR and SS-SBR selected microbial consortia with markedly different surfactant-producing and PAH-degrading abilities. Biosurfactant levels in the SS-SBR reached 4 times the critical micelle concentration (CMC) that resulted in considerable emulsification of PAH. In contrast, CSTR operation resulted in nomeasurable biosurfactant production. Total PAH removal efficiency was 93% in the SS-SBR, compared with only 66% in the CSTR, and stripping of PAH was 3 times less in the SS-SBR. Reversing the mode of operation on day 100 caused a complete reversal in microbial consortia and in reactor performance by day 140. These results show that bioslurry reactor operation can be manipulated to control overall reactor performance.

  12. Global physical and numerical stability of a nuclear reactor core

    International Nuclear Information System (INIS)

    Morales-Sandoval, Jaime; Hernandez-Solis, Augusto

    2005-01-01

    Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented

  13. Opportunities for physics research at Australia's replacement research reactor

    International Nuclear Information System (INIS)

    Robinson, R.A.

    2003-01-01

    Full text: The 20-MW Australian Replacement Research Reactor represents possibly the greatest single research infrastructure investment in Australia's history. Construction of the facility has commenced, following award of the construction contract in July 2000, and the construction licence in April 2002. The project includes a large state-of-the-art liquid deuterium cold-neutron source and supermirror guides feeding a large modern guide hall, in which most of the instruments are placed. Alongside the guide hall, there is good provision of laboratory, office and space for support activities. While the facility has 'space' for up to 18 instruments, the project has funding for an initial set of 8 instruments, which will be ready when the reactor is fully operational in January 2006. Instrument performance will be competitive with the best research-reactor facilities anywhere, and our goal is to be in the top 3 such facilities worldwide. Staff to lead the design effort and man these instruments have been hired on the international market from leading overseas facilities, and from within Australia, and 6 out of 8 instruments have been specified and costed. At present the instrumentation project carries ∼15% contingency. An extensive dialogue has taken place with the domestic user community and our international peers, via various means including a series of workshops over the last 2 years covering all 8 instruments, emerging areas of application like biology and the earth sciences, and computing infrastructure for the instruments. In December 2002, ANSTO formed the Bragg Institute, with the intent of nurturing strong external partnerships, and covering all aspects of neutron and X-ray scattering, including research using synchrotron radiation. I will discuss the present status and predicted performance of the neutron-beam facilities at the Replacement Reactor, synergies with the synchrotron in Victoria, in-house x-ray facilities that we intend to install in the Bragg

  14. Educational reactor-physics experiments with the critical assemble TCA

    Energy Technology Data Exchange (ETDEWEB)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

    1997-10-01

    The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

  15. Methane production by treating vinasses from hydrous ethanol using a modified UASB reactor.

    Science.gov (United States)

    España-Gamboa, Elda I; Mijangos-Cortés, Javier O; Hernández-Zárate, Galdy; Maldonado, Jorge A Domínguez; Alzate-Gaviria, Liliana M

    2012-11-21

    A modified laboratory-scale upflow anaerobic sludge blanket (UASB) reactor was used to obtain methane by treating hydrous ethanol vinasse. Vinasses or stillage are waste materials with high organic loads, and a complex composition resulting from the process of alcohol distillation. They must initially be treated with anaerobic processes due to their high organic loads. Vinasses can be considered multipurpose waste for energy recovery and once treated they can be used in agriculture without the risk of polluting soil, underground water or crops. In this sense, treatment of vinasse combines the elimination of organic waste with the formation of methane. Biogas is considered as a promising renewable energy source. The aim of this study was to determine the optimum organic loading rate for operating a modified UASB reactor to treat vinasse generated in the production of hydrous ethanol from sugar cane molasses. The study showed that chemical oxygen demand (COD) removal efficiency was 69% at an optimum organic loading rate (OLR) of 17.05 kg COD/m3-day, achieving a methane yield of 0.263 m3/kg CODadded and a biogas methane content of 84%. During this stage, effluent characterization presented lower values than the vinasse, except for potassium, sulfide and ammonia nitrogen. On the other hand, primers used to amplify the 16S-rDNA genes for the domains Archaea and Bacteria showed the presence of microorganisms which favor methane production at the optimum organic loading rate. The modified UASB reactor proposed in this study provided a successful treatment of the vinasse obtained from hydrous ethanol production.Methanogen groups (Methanobacteriales and Methanosarcinales) detected by PCR during operational optimum OLR of the modified UASB reactor, favored methane production.

  16. Methane production by treating vinasses from hydrous ethanol using a modified UASB reactor

    Directory of Open Access Journals (Sweden)

    España-Gamboa Elda I

    2012-11-01

    Full Text Available Abstract Background A modified laboratory-scale upflow anaerobic sludge blanket (UASB reactor was used to obtain methane by treating hydrous ethanol vinasse. Vinasses or stillage are waste materials with high organic loads, and a complex composition resulting from the process of alcohol distillation. They must initially be treated with anaerobic processes due to their high organic loads. Vinasses can be considered multipurpose waste for energy recovery and once treated they can be used in agriculture without the risk of polluting soil, underground water or crops. In this sense, treatment of vinasse combines the elimination of organic waste with the formation of methane. Biogas is considered as a promising renewable energy source. The aim of this study was to determine the optimum organic loading rate for operating a modified UASB reactor to treat vinasse generated in the production of hydrous ethanol from sugar cane molasses. Results The study showed that chemical oxygen demand (COD removal efficiency was 69% at an optimum organic loading rate (OLR of 17.05 kg COD/m3-day, achieving a methane yield of 0.263 m3/kg CODadded and a biogas methane content of 84%. During this stage, effluent characterization presented lower values than the vinasse, except for potassium, sulfide and ammonia nitrogen. On the other hand, primers used to amplify the 16S-rDNA genes for the domains Archaea and Bacteria showed the presence of microorganisms which favor methane production at the optimum organic loading rate. Conclusions The modified UASB reactor proposed in this study provided a successful treatment of the vinasse obtained from hydrous ethanol production. Methanogen groups (Methanobacteriales and Methanosarcinales detected by PCR during operational optimum OLR of the modified UASB reactor, favored methane production.

  17. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR

    International Nuclear Information System (INIS)

    CHENG, L.; HANSON, A.; DIAMOND, D.; XU, J.; CAREW, J.; RORER, D.

    2004-01-01

    Detailed reactor physics and safety analyses have been performed for the 20 MW D 2 O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the core

  18. Physical aspects of liquid-impelled loop reactors

    NARCIS (Netherlands)

    Sonsbeek, van H.

    1992-01-01

    The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the

  19. Job analysis of nuclear power reactor health physics technicians

    International Nuclear Information System (INIS)

    Davis, L.T.; Mazour, T.J.; Clark, P.V.; Todd, R.C.; Marotta, F.J.

    1984-06-01

    This report describes a project, an industry-wide Job Analysis of Nuclear Power Reactor Health Physics Technicians (HPTs), conducted by Brookhaven National Laboratory and Analysis and Technology, Inc. to provide the industry with job-performance data that can be used in systematically defining training programs in terms of required job functions responsibilities, and performance standards. The job-analysis methodology is consistent with that used by the Institute of Nuclear Power Operations (INPO) in similar industry-wide projects and includes administration of over 850 job task questionnaires to utility and contractor Health Physics Technicians throughout the country. Data collected includes task performance (difficulty, importance, and frequency) and industry-wide demographics (job levels, experience, education, and training). The results of this project discussed herein include model job descriptions for HPT positions, summaries of HPT experience, education, and training, industry-wide task listings with task-performance characteristics, and recommendations of selected tasks as a basis for HPT training development. Finally, potential future applications of the data base by utility and contractor organizations in training program development and evaluation and personnel qualifications are discussed

  20. PERFORMANCE EVALUATION OF AN ANAEROBIC BAFFLED REACTOR TREATING WHEAT FLOUR STARCH INDUSTRY WASTEWATER

    Directory of Open Access Journals (Sweden)

    H. Movahedyan, A. Assadi, A. Parvaresh

    2007-04-01

    Full Text Available Feasibility of the anaerobic baffled reactor process was investigated for the treatment of wheat flour starch wastewater. After removal of suspended solids by simple gravity settling, starch wastewater was used as a feed. Start-up of a reactor (with a volume of 13.5 L and five compartments with diluted feed of approximately 4500 mg/L chemical oxygen demand was accomplished in about 9 weeks using seed sludge from anaerobic digester of municipal wastewater treatment plant. The reactor with hydraulic retention time of 72h at 35°C and initial organic loading rate of 1.2 kgCOD/m3.d showed 61% COD removal efficiency. The best performance of reactor was observed with an organic loading rate of 2.5 kgCOD/m3.d or hydraulic retention time of 2.45 d and the COD conversion of 67% was achieved. The system also showed very high solids retention with effluent suspended solids concentration of about 50 mg/L for most organic and hydraulic loadings studied. Based on these observations, the ABR process has potential to treat food industrial wastewater as a pretreatment and is applicable for extreme environmental conditions.

  1. Opportunities for applied measurements using the PROSPECT antineutrino detector: reactor physics and safeguards

    Science.gov (United States)

    Bowden, Nathaniel; Prospect Collaboration

    2015-10-01

    Disagreement of reactor antineutrino spectrum and flux measurements with updated predictions indicates that we have much to learn about the complicated processes underlying antineutrino production in reactors, as well as hinting at new physics. A number of new efforts seek to address these questions, including the PROSPECT experiment planned at the HFIR research reactor. In addition to greatly advancing our understanding of reactor antineutrino emissions, PROSPECT can support a rich applied physics program. The detection technology developed for PROSPECT will enable precision antineutrino spectrum measurements close to essentially any reactor type. Here we describe how such measurements provide opportunities to probe fissile isotope and fission daughter distributions, and their potential use for reactor physics and safeguards applications. LLNL-ABS-673983. Prepared by LLNL under Contract DE-AC52-07NA27344.

  2. The mechanics in the reactors physics; La mecanique dans la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Dept. d' Etudes des Reacteurs, DER/SPRC, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee, DRFC, 13 - Saint-Paul-lez-Durance (France)] [and others

    1998-12-22

    This meeting of the 24 november 1998, took place in Paris and was organized by the SFEN. After three plenary sessions a technical meeting dealt on the mechanics in reactors physics. The plenary papers presented the state of the art in the PWR type reactors and fast neutron reactors systems and in the thermonuclear reactors system. Five more technical papers presented the seismic behavior of the reactors cores, the fuel-cladding interactions, the defects harmfulness in the fracture mechanics and the fuel rods control system wear. (A.L.B.)

  3. A modular diagnosis system based on fuzzy logic for UASB reactors treating sewage.

    Science.gov (United States)

    Borges, R M; Mattedi, A; Munaro, C J; Franci Gonçalves, R

    A modular diagnosis system (MDS), based on the framework of fuzzy logic, is proposed for upflow anaerobic sludge blanket (UASB) reactors treating sewage. In module 1, turbidity and rainfall information are used to estimate the influent organic content. In module 2, a dynamic fuzzy model is used to estimate the current biogas production from on-line measured variables, such as daily average temperature and the previous biogas flow rate, as well as the organic load. Finally, in module 3, all the information above and the residual value between the measured and estimated biogas production are used to provide diagnostic information about the operation status of the plant. The MDS was validated through its application to two pilot UASB reactors and the results showed that the tool can provide useful diagnoses to avoid plant failures.

  4. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and The International Reactor Reactor Physics Experiment Evaluation Project (IRPhEP)

    Energy Technology Data Exchange (ETDEWEB)

    Briggs, J.B.; Bess, J. [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Gulliford, J. [Organization for Economic Cooperation and Development (OECD),Nuclear Energy Agency, Paris, (France)

    2011-07-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) are sources of evaluated integral benchmark data that may be used for validation of reactor physics / nuclear criticality safety analytical methods and data, nuclear data testing, and safety analysis licensing activities. The IRPhEP is patterned after its predecessor, the ICSBEP, but focuses on other integral measurements such as buckling, spectral characteristics, reactivity effects, reactivity coefficients, kinetics measurements, reaction-rate and power distributions, nuclide compositions and other miscellaneous types of measurements in addition to the critical configuration. Both projects will be discussed.

  5. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  6. Present status of reactor physics in the United States and Japan-IV. 2. Micro-Reactor Physics of MOX-Fueled Core

    International Nuclear Information System (INIS)

    Takeda, Toshikazu

    2001-01-01

    Recently, fuel assemblies of light water reactors have become complicated because of the extension of fuel burnup and the use of high-enriched Gd and mixed-oxide (MOX) fuel, etc. In conventional assembly calculations, the detailed flux distribution, spectrum distribution, and space dependence of self-shielding within a fuel pellet are not directly taken into account. The experimental and theoretical study of investigating these microscopic properties is named micro-reactor physics. The purpose of this work is to show the importance of micro-reactor physics in the analysis of MOX fuel assemblies. Several authors have done related studies; however, their studies are limited to fuel pin cells, and they are never mentioned with regard to burnup effect, which is important for actual core design. We used the subgroup method to treat the space dependence of the self-shielding effect of heavy nuclides, and we used the characteristics method to treat the angular dependence of neutron flux in a fuel pellet. Figure 1 compares the power distributions in MOX and UO 2 fuel cells at the beginning of burnup. The power is calculated with and without considering the space dependence of the self-shielding effect of the cross sections. For the MOX cell, the power distribution has a peak at the cell edge because of large Pu absorption especially when considering the spatial self-shielding effect. When a MOX rod is adjacent to UO 2 fuel rods, the flux distribution has an azimuthal dependence in addition to the radial dependence within a rod. For example, consider a 2x2 fuel assembly composed of three UO 2 rods and one MOX rod, with the mirror reflection boundary condition. A burnup calculation was done with the condition; the radius of the MOX pellet is divided into two regions, and the azimuthal angle is divided into eight. The number density of 239 Pu at 44 000 MWd/t for the MOX rod shows azimuthal dependence by 20%. The maximum burnup occurs in the direction of the UO 2 rods. This is

  7. Health physics aspects of advanced reactor licensing reviews

    Energy Technology Data Exchange (ETDEWEB)

    Hinson, C.S. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-03-01

    The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.

  8. Modelling of thermalhydraulics and reactor physics in simulators

    International Nuclear Information System (INIS)

    Miettinen, J.

    1994-01-01

    The evolution of thermalhydraulic analysis methods for analysis and simulator purposes has brought closer the thermohydraulic models in both application areas. In large analysis codes like RELAP5, TRAC, CATHARE and ATHLET the accuracy for calculating complicated phenomena has been emphasized, but in spite of large development efforts many generic problems remain unsolved. For simulator purposes fast running codes have been developed and these include only limited assessment efforts. But these codes have more simulator friendly features than large codes, like portability and modular code structure. In this respect the simulator experiences with SMABRE code are discussed. Both large analysis codes and special simulator codes have their advances in simulator applications. The evolution of reactor physical calculation methods in simulator applications has started from simple point kinetic models. For analysis purposes accurate 1-D and 3-D codes have been developed being capable for fast and complicated transients. For simulator purposes capability for simulation of instruments has been emphasized, but the dynamic simulation capability has been less significant. The approaches for 3-dimensionality in simulators requires still quite much development, before the analysis accuracy is reached. (orig.) (8 refs., 2 figs., 2 tabs.)

  9. Analysis of deficiencies in fast reactor blanket physics predictions

    International Nuclear Information System (INIS)

    Hill, R.N.

    1987-12-01

    This analysis addresses a deviation between experimental measurements and fast reactor blanket physics predictions. A review of worldwide results reveals that reaction rates in the blanket are underpredicted with the discrepancy increasing with penetration into the blanket. The analysis of this discrepancy involves two parts: quantifying possible error reductions using the most advanced methods and investigating deficiencies in current methodology. The source of these discrepancies was investigated by application of ''state-of-the-art'' group constant generation and flux prediction methodology to flux calculations for the Purdue University Fast Breeder Blanket Facility (FBBF). Refined group constant generation methods yielded a significant reduction in the blanket deviations; however, only about half of the discrepancy can be accounted for in this manner. Transport theory calculations were used to predict the blanket neutron transmission problem. The surprising result is that transport theory predictions utilizing diffusion theory group constants did not improve the blanket results. Transport theory predictions exhibited blanket underpredictions similar to the diffusion theory results. The residual blanket discrepancies not explained using advanced methods require a refinement of the theory. For this purpose an analysis of deficiencies in current methodology was performed

  10. Estimation of optimum operating parameters of UASB reactor treating flax retting wastewater by kinetic model

    Energy Technology Data Exchange (ETDEWEB)

    Liu, J.; Ukita, M.; Nakanishi, H.; Imai, T. [Yamaguchi University, Yamaguchi (Japan); Fukagawa, M. [Ube Technical College, Yamaguchi (Japan)

    1995-08-21

    A laboratory study was used to develop a simplified kinetic model, to evaluate the kinetic parameters, and to provide rational design parameters for a pilot plant treating flax retting wastewater by means of the simulation of optimal operation of the UASB reactor. The results indicated that the developed model can be used predicatively for assessing plant performance and when the concentration of the influent is at the range of 5.5-7.3gCOD/l, the concentration of the hard-biodegradable materials is 0.46 gCOD/l. 14 refs., 9 figs., 3 tabs.

  11. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and

  12. Occurence of methanogenesis during start-up of a full-scale synthesis gas-fed reactor treating sulfate and metal-rich wastewater

    NARCIS (Netherlands)

    Houten, van B.H.G.W.; Roest, C.; Tzeneva, V.A.; Dijkman, H.; Smidt, H.; Stams, A.J.M.

    2006-01-01

    The start-up of a full-scale synthesis gas-fed gas-lift reactor treating metal and sulfate-rich wastewater was investigated. Sludge from a pilot-scale reactor was used to seed the full-scale reactor. The main difference in design between the pilot- and full-scale reactor was that metal precipitation

  13. Future view of total energy system and reactor engineering and reactor physics

    International Nuclear Information System (INIS)

    Ozawa, T.

    1974-01-01

    This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

  14. 78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2013-08-19

    ... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Orders; rescission. SUMMARY... the NRC published a final rule, ``Physical Protection of Irradiated Fuel in Transit,'' on May 20, 2013... of Irradiated Reactor Fuel in Transit'' (RIN 3150-AI64; NRC-2009-0163). The final rule incorporates...

  15. 78 FR 29519 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2013-05-20

    ... Protection of Irradiated Reactor Fuel in Transit; Final Rule #0;#0;Federal Register / Vol. 78 , No. 97... Part 73 RIN 3150-AI64 [NRC-2009-0163] Physical Protection of Irradiated Reactor Fuel in Transit AGENCY... orders for SNF in transit? F. When will the NRC issue guidance on these requirements? G. What is...

  16. The development of the physical conceptions of the FBR type reactors control methods

    International Nuclear Information System (INIS)

    Matveev, V.I.; Ivanov, A.P.

    1984-01-01

    The physical concepts and specific problems of the control elements for LMFBR type reactors are discussed in this paper. Typical temperature coefficient of reactivity, its dependency on reactor power and burnup level are given. The authors give us the most advisable methods of the reactivity coefficient compensation

  17. Inspection methods for physical protection Task III review of other agencies' physical security activities for research reactors

    International Nuclear Information System (INIS)

    In Task I of this project, the current Nuclear Regulatory Commission (NRC) position-on physical security practices and procedures at research reactors were reviewed. In the second task, a sampling of the physical security plans was presented and the three actual reactor sites described in the security plans were visited. The purpose of Task III is to review other agencies' physical security activities for research reactors. During this phase, the actions, procedures and policies of two domestic and two foreign agencies other than the NRC that relate to the research reactor community were examined. The agencies examined were: International Atomic Energy Agency; Canadian Atomic Energy Control Board; Department of Energy; and American Nuclear Insurers

  18. Successional development of biofilms in moving bed biofilm reactor (MBBR) systems treating municipal wastewater.

    Science.gov (United States)

    Biswas, Kristi; Taylor, Michael W; Turner, Susan J

    2014-02-01

    Biofilm-based technologies, such as moving bed biofilm reactor (MBBR) systems, are widely used to treat wastewater. Biofilm development is important for MBBR systems as much of the microbial biomass is retained within reactors as biofilm on suspended carriers. Little is known about this process of biofilm development and the microorganisms upon which MBBRs rely. We documented successional changes in microbial communities as biofilms established in two full-scale MBBR systems treating municipal wastewater over two seasons. 16S rRNA gene-targeted pyrosequencing and clone libraries were used to describe microbial communities. These data indicate a successional process that commences with the establishment of an aerobic community dominated by Gammaproteobacteria (up to 52 % of sequences). Over time, this community shifts towards dominance by putatively anaerobic organisms including Deltaproteobacteria and Clostridiales. Significant differences were observed between the two wastewater treatment plants (WWTPs), mostly due to a large number of sequences (up to 55 %) representing Epsilonproteobacteria (mostly Arcobacter) at one site. Archaea in young biofilms included several lineages of Euryarchaeota and Crenarchaeota. In contrast, the mature biofilm consisted entirely of Methanosarcinaceae (Euryarchaeota). This study provides new insights into the community structure of developing biofilms at full-scale WWTPs and provides the basis for optimizing MBBR start-up and operational parameters.

  19. Proceedings of the symposium on the physics and technology of reactors

    International Nuclear Information System (INIS)

    1993-01-01

    The symposium aimed at providing the opportunity for promoting the subject and for developing the human resources in this important field in the Arab States. The symposium included 32 lectures on the following topics related to research reactors: design and development, training and operation, calculations of reactor parameters, nuclear reactions dynamics and control, reactor physics, neutron pyhsics, neutron activation analysis, in-core reactor radiation protection and shielding calculations. The lectures of the symposium were distributed over 7 sessions. An additional session was held by all participants for open discussion and recommendations

  20. Design of data sampler in intelligent physical start-up system for nuclear reactor

    International Nuclear Information System (INIS)

    Wang Yinli; Ling Qiu

    2007-01-01

    It introduces the design of data sampler in intelligent physical start-up system for nuclear reactor. The hardware frame taking STμPSD3234A as the core and the firmware design based on USB interface are discussed. (authors)

  1. Reactor physics activities in NEA member countries October 1990-September 1991

    International Nuclear Information System (INIS)

    1991-01-01

    This document is a compilation of National Activity Reports presented at the Thirty-Fourth Meeting of the NEA Committee on Reactor Physics, held at the Paul Scherrer Institute, Wuerenlingen, Switzerland, from 3rd-5th September 1991

  2. A WIMS-NESTLE reactor physics model for an RBMK reactor

    International Nuclear Information System (INIS)

    Perry, R.T.; Meriwether, G.H.

    1996-01-01

    This work describes the static neutronic calculations made for a three-dimensional model of an RBMK (Russian) reactor. Future work will involve the use of this neutronic model and a thermal-hydraulic model in coupled calculations. The lattice code, WIMS-D, was used to obtain the cross sections for the static neutronic calculations. The static reactor neutronic calculations were made with NESTLE, a three-dimensional nodal diffusion code. The methods used to establish an RBMK reactor model for use in these codes are discussed, and the cross sections calculated are given

  3. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    International Nuclear Information System (INIS)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-01-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean/US/laboratory/university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program

  4. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  5. Seismic hazard study for the TREAT Reactor facility at the INEL, Idaho

    International Nuclear Information System (INIS)

    1979-01-01

    The TREAT Reactor is founded on a thick unfaulted sequence of Plio-Pleistocene basalt on the Snake River Plain. The plain is presently aseismic; however, seismic activity occurs in the mountains around the plain. The Howe Scarp is located 19 miles from TREAT and contains a known capable fault. Evaluation of this and other faults in the region indicate the Howe Scarp is the most significant earthquake fault for TREAT. A maximum credible earthquake on this fault could produce a maximum ground motion of about .22 g at TREAT. A study of three range front fault systems north of the Snake River Plain indicates the fault systems have not ruptured as a unit in the past; and, cross range faults, mountain spurs and reentrants generally bound the definable fault sets in the range front systems. This study indicates future surface fault rupture and earthquake events will follow a similar pattern of contiguous faulting; each individual surface rupture event should only involve a single fault set of the range front fault system. Surface faulting on contiguous fault sets should be separated by significant intervals of geologic time. Certain volcanic hazards have been examined and discussed

  6. Benchmarking lattice physics data and methods for boiling water reactor analysis

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Edenius, M.; Harris, D.R.; Hebert, M.J.; Kapitz, D.M.; Pilat, E.E.; VerPlanck, D.M.

    1983-01-01

    The objective of the work reported was to verify the adequacy of lattice physics modeling for the analysis of the Vermont Yankee BWR using a multigroup, two-dimensional transport theory code. The BWR lattice physics methods have been benchmarked against reactor physics experiments, higher order calculations, and actual operating data

  7. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Science.gov (United States)

    2010-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1...

  8. Impact of confinement physics on reactor design and economics

    International Nuclear Information System (INIS)

    DeFreece, D.A.; Campbell, R.B.; Waganer, L.M.

    1977-01-01

    A variety of confinement laws were employed in a transient, zero dimensional plasma code, which was coupled to the TOCOMO systems code. The purpose was to determine the impact of the confinement laws on reactor design, power costs and changes in the utility interface. A satisfactory reactor and power plant has been defined for the large majority of combinations of confinement law, power plant size and plasma shape. Trapped ion mode (TIM) has been the easiest to work with, since the plasma is thermally stable with a good power density and minimal alpha particle build up. Neoclassical and pseudoclassical along with TEMII result in satisfactory reactor performance, but require active feedback control (by injecting impurities) to prevent plasma temperature excursions. These laws also require some form and degree of confinement time spoiling to allow long burn times, otherwise, alpha particles build up to an unacceptable level. TEM I results in thermal equilibrium at 5 keV and must be driven to provide a reactor quality plasma. The continuous injected power required for a 4300 MW thermal reactor is 540 MW. This added to the other circulating loads results in a net power output of 600 MWe at a very high relative cost. Daughney (empirical) confinement results in a satisfactory, competitive reactor

  9. Test on the reactor with the portable digital reactivity meter for physical experiment

    International Nuclear Information System (INIS)

    Huang Liyuan

    2010-01-01

    Test must be performed on the zero power reactor During the development of portable digital reactivity meter for physical experiment, in order to check its measurement function and accuracy. It describes the test facility, test core, test methods, test items and test results. The test results show that the instrument satisfy the requirements of technical specification, and satisfy the reactivity measurement in the physical experiments on reactors. (authors)

  10. Study and application of digital physical start-up system for nuclear reactor

    International Nuclear Information System (INIS)

    Qu Ronghong; Li Baoxiang; Xu Xiaolin

    2004-01-01

    The digital physical start-up system for nuclear reactor is introduced. The system was used successfully in physical start-up experiment of 10 MW high-temperature gas-cooled reactor. It is proved practically that the system not only runs reliably and calculates both rapidly and correctly and relieves the loads of operators, but also has the better characters of monitoring and showing the real-time results of experiments than the analog systems. (author)

  11. Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1989-06-01

    The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

  12. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  13. Fusion reactor physics and technology. Progress report, October 1, 1978-June 30, 1979

    International Nuclear Information System (INIS)

    Conn, R.W.; Kulcinski, G.L.; Maynard, C.W.

    1979-01-01

    During the present contract period, work has been carried out in the following areas: (a) The NUWMAK tokamak reactor design was completed and distributed throughout the community. In particular, specific work was completed on divertorless tokamak operation in NUWMAK, Ti alloy assessment, materials resource implications of NUWMAK style reactors, and an economic analysis; (b) Tandem mirror reactor technology studies were carried out on tandem mirror physics, the role of rf heating, power balance studies, the design of high field magnets, and blanket/shield design in TMR's; (c) work at Wisconsin is contributing to the evolving picture of an optimum TMR; (d) the WHIST tokamak reactor plasma transport code developed at Wisconsin has been extended in two directions; (e) Work on ICRF heating in tokamak reactors, both in terms of physics and launching structure design, has been completed and published

  14. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  15. New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford

    2012-11-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.

  16. Photocatalytic reactors for treating water pollution with solar illumination: a simplified analysis for n-steps flow reactors with recirculation

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Universitaet Hannover (Germany). Institut fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [INTEC Universidad Nacional del Litoral and CONICET, Sante Fe (Argentina)

    2005-09-01

    The concentration of dissolved oxygen in water, in equilibrium with atmospheric air (ca. 8 ppm at 20{sup o}C), defines the limits of all practical oxidizing processes for removing pollutants in photocatalytic reactors. To solve this limitation, an alternative approach to that of a continuously aerated reactor is the use of a recirculating system with aeration performed after every cycle at the reactor entering stream. As defined by the nature of a single recirculating step (the need of a reactor operation at a rather low concentration range), this procedure results in a very low photonic efficiency (thus requiring a large photon collecting area and consequently increasing the capital cost). The design engineer will have to resort to a series of several reactors with recirculation. This solution may then lead to a very high Photonic Efficiency for the entire process (i.e., a reduced light harvesting area) at the price of an increase in the required capital cost (due to the larger number of reactors). This paper provides a very simple analysis and analytical expressions that can be used to estimate, for a desired degree of degradation, a trade-off solution between a high number of reactors and a very large surface area to collect the solar photons. (author)

  17. Reactor physics standards: The key to successfully dealing with technical diversity

    International Nuclear Information System (INIS)

    Knuckles, E.R.

    1990-01-01

    Effective and valuable reactor physics standards can be successfully developed to accommodate diversity in available calculations tools and promote improvement in existing methods. The issues encountered and lessons learned in the standard, 'Calculation of Doppler Reactivity for Use in Thermal Light Water Reactor Analysis' (now under development by the ANS 19.7 working group), demonstrate this point. With a diversity of reactor physics tools available, differing levels of user experience, and a variety of procedures for calculating reactor physics parameters important to safety, it is not surprising that there are differing levels of quality in the calculational result of the same parameters. An approach that effectively deals with this technical diversity is standardization of the calculational process. This approach assures that the user's expectations are consistently met. In order for a standard to be effective, it must recognize and address three essential elements of this process: the user, a set of codes, and associated procedures

  18. Performance of plastic- and sponge-based trickling filters treating effluents from an UASB reactor.

    Science.gov (United States)

    Almeida, P G S; Marcus, A K; Rittmann, B E; Chernicharo, C A L

    2013-01-01

    The paper compares the performance of two trickling filters (TFs) filled with plastic- or sponge-based packing media treating the effluent from an upflow anaerobic sludge blanket (UASB) reactor. The UASB reactor was operated with an organic loading rate (OLR) of 1.2 kgCOD m(-3) d(-1), and the OLR applied to the TFs was 0.30-0.65 kgCOD m(-3) d(-1) (COD: chemical oxygen demand). The sponge-based packing medium (Rotosponge) gave substantially better performance for ammonia, total-N, and organic matter removal. The superior TF-Rotosponge performance for NH(4)(+)-N removal (80-95%) can be attributed to its longer biomass and hydraulic retention times (SRT and HRT), as well as enhancements in oxygen mass transfer by dispersion and advection inside the sponges. Nitrogen removals were significant (15 mgN L(-1)) in TF-Rotosponge when the OLRs were close to 0.75 kgCOD m(-3) d(-1), due to denitrification that was related to solids hydrolysis in the sponge interstices. For biochemical oxygen demand removal, higher HRT and SRT were especially important because the UASB removed most of the readily biodegradable organic matter. The new configuration of the sponge-based packing medium called Rotosponge can enhance the feasibility of scaling-up the UASB/TF treatment, including when retrofitting is necessary.

  19. Sludge accumulation in shallow maturation ponds treating UASB reactor effluent: results after 11 years of operation.

    Science.gov (United States)

    Possmoser-Nascimento, Thiago Emanuel; Rodrigues, Valéria Antônia Justino; von Sperling, Marcos; Vasel, Jean-Luc

    2014-01-01

    Polishing ponds are natural systems used for the post-treatment of upflow anaerobic sludge blanket (UASB) effluents. They are designed as maturation ponds and their main goal is the removal of pathogens and nitrogen and an additional removal of residual organic matter from the UASB reactor. This study aimed to evaluate organic matter and suspended solids removal as well as sludge accumulation in two shallow polishing ponds in series treating sanitary effluent from a UASB reactor with a population equivalent of 200 inhabitants in Brazil, operating since 2002. For this evaluation, long-term monitoring of biochemical oxygen demand and total suspended solids and bathymetric surveys have been undertaken. The ponds showed an irregular distribution of total solids mass in the sludge layer of the two ponds, with mean accumulation values of 0.020 m(3) person(-1) year(-1) and 0.004 m(3) person(-1) year(-1) in Ponds 1 and 2, leading to around 40% and 8% of the liquid volume occupied by the sediments after 11 years of operation. The first pond showed better efficiency in relation to organic matter removal, although its contribution was limited, due to algal growth. No simple input-output mass balance of solids can be applied to the ponds due to algal growth in the liquid phase and sludge digestion in the sludge.

  20. Treating the neutronics of a (d,t)-fusion reactor operating on the tokamak principle (NET)

    International Nuclear Information System (INIS)

    Fischer, U.

    1990-10-01

    The primary aim of this work is to check the simplified one-dimensional neutronic approach being used frequently in design calculations. Therefore the neutronics of the NET (Next European Torus)-reactor is treated in a three-dimensional torus sector model by means of Monte Carlo calculations with the MCNP-code. Various blanket variants with different neutronic characteristics are taken into account. All of them had been developed for use in the NET-reactor: a ceramic solid breeder blanket using beryllium as neutron multiplier, a self-cooled liquid metal blanket using the eutectic alloy Pb-17Li or, alternatively, pure lithium as breeding material/coolant, and an aqueous lithium salt solution blanket. It is shown, that the one-dimensional approach can be applied in design calculations for evaluating power density distributions, if the plasma source is normalized in a consistent manner, if its spatial distribution is choosen appropriately, and, furthermore, if its angular dependence is taken into account. A three-dimensional treatment of the actual tokamak geometry, however, is necessary for determining the breeding performance of the blanket variants and for performing shielding calculations of the whole system blanket/shield. (orig./HP) [de

  1. Efficiency of a Bed Biofilm Reactor Using a LECA Carrier to Treat Hospital Wastewater

    Directory of Open Access Journals (Sweden)

    Reza Shokoohi

    2016-03-01

    Full Text Available Hospital wastewater is of great environmental concern because it contains a variety of hazardous microbial and chemical substances. This study aims to investigate the efficiency of a moving bed biofilm reactor (MBBR with a light expanded clay aggregate (LECA carrier for treating hospital wastewater. This pilot study used a Plexiglas reactor that had a volume of 100 L, a continuous up-flow hydraulic regime, and a LECA carrier to test removal of chemical oxygen demand (COD from wastewater in a public hospital. To assess MBBR efficiency, the study used retention times of 8, 12, and 24 hours, filling percentages of 30% and 50%, and mixed liquor suspended solids (MLSSs of 1000, 3000, and 5000 mg/L. The results indicated that using a single LECA carrier in an MBBR system was not sufficient to remove organic materials from hospital wastewater, because the carrier could not be completely suspended. After some modifications, consisting mainly of returning activated sludge, the system was 83% efficient at removing COD using a LECA carrier at a retention time of 24 hours, with 50% filling, and 5000 mg/L of MLSS.

  2. The under-critical reactors physics for the hybrid systems

    International Nuclear Information System (INIS)

    Schapira, J.P.; Vergnes, J.; Zaetta, A.

    1998-01-01

    This day, organized by the SFEN, took place at Paris the 12 march 1998. Nine papers were presented. They take stock on the hybrid systems and more specifically the under-critical reactors. One of the major current preoccupation of nuclear industry is the problems of the increase of radioactive wastes produced in the plants and the destruction of the present stocks. To solve these problems a solution is the utilisation of hybrid systems: the coupling of a particle acceleration to an under-critical reactor. Historical aspects, advantages and performances of such hybrid reactors are presented in general papers. More technical papers are devoted to the spallation, the MUSE and the TARC experiments. (A.L.B.)

  3. Performance evaluation of full scale UASB reactor in treating stillage wastewater

    OpenAIRE

    A.Mirsepasi , H. R. Honary , A. R. Mesdaghinia, A. H. Mahvi , H. Vahid , H. Karyab

    2006-01-01

    Upflow anaerobic sludge blanket (UASB) reactors have been widely used for treatment of industrial wastewater. In this study two full-scale UASB reactors were investigated. Volume of each reactor was 420 m3. Conventional parameters such as pH, temperature and efficiency of COD, BOD, TOC removal in each reactor were investigated. Also several initial parameters in designing and operating of UASB reactors, such as upflow velocity, organic loading rate (OLR) and hydraulic retention time were inve...

  4. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  5. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    International Nuclear Information System (INIS)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched 235 U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched 235 U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing

  6. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joseph W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Norman, Daren R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-01

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

  7. Fast Reactor Physics Vol. I. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  8. Possible physics modifications to CIRUS reactor core for improved reactor utilization

    International Nuclear Information System (INIS)

    John, Benjamin; Khosla, S.K.; Narain, Rajendra.

    1976-01-01

    Two fuelling schemes for uprating the neutron flux in CIRUS reactor at Trombay, are studied. One scheme employs enriched uranium-aluminium alloy boosters, the second envisages employing thorium oxide enriched with 0.2% plutonium oxide. It is seen that the second scheme has the potential of in-situ thorium utilization. (M.G.B.)

  9. Evaluation of a hybrid anaerobic biofilm reactor treating winery effluents and using grape stalks as biofilm carrier.

    Science.gov (United States)

    Wahab, Mohamed Ali; Habouzit, Frédéric; Bernet, Nicolas; Jedidi, Naceur; Escudié, Renaud

    2016-01-01

    Wine production processes generate large amount of both winery wastewater and solid wastes. Furthermore, working periods, volumes and pollution loads greatly vary over the year. Therefore, it is recommended to develop a low-cost treatment technology for the treatment of winery effluents taking into account the variation of the organic loading rate (OLR). Accordingly, we have investigated the sequential operation of an anaerobic biofilm reactor treating winery effluents and using grape stalks (GSs) as biofilm carrier with an OLR ranging from 0.65 to 27 gCOD/L/d. The result showed that, during the start-up with wastewater influent, the chemical oxygen demand (COD) removal rate ranged from 83% to 93% and was about 91% at the end of the start-up period that lasted for 40 days. After 3 months of inactivity period of the reactor (no influent feeding), we have succeeded in restarting-up the reactor in only 15 days with a COD removal of 82% and a low concentration of volatile fatty acids (1 g/L), which confirms the robustness of the reactor. As a consequence, GSs can be used as an efficient carrier support, allowing a fast reactor start-up, while the biofilm conserves its activity during a non-feeding period. The proposed hybrid reactor thus permits to treat both winery effluents and GSs.

  10. High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

    Science.gov (United States)

    Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

    2014-01-01

    An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

  11. Franco-German cooperation for the physical protection of the EPR reactor

    International Nuclear Information System (INIS)

    Jalouneix, J.; Hagemann, A.

    2001-01-01

    This article presents the proceeding that has been followed in the EPR (European pressurized water reactor) project concerning physical protection against malevolent actions and robbery of nuclear materials. Before the different options of the nuclear island were definitely set, a task group had been constituted to examine if these options could hamper the setting of physical protection measures that are required by the legislation of the 2 countries. Another group composed of experts from IPSN/GRS (Institut de Protection et de Surete Nucleaire / Gesellschaft fur Anlagen und Reaktorsicherheit) had the task to define common requirements concerning the physical protection of reactors in Germany and in France. In this framework the EPR project team has prepared a technical document reviewing the different dispositions that have been retained to assure the physical protection of the reactor. (A.C.)

  12. Coupled hydrodynamic-structural analysis of an integral flowing sodium test loop in the TREAT reactor

    International Nuclear Information System (INIS)

    Zeuch, W.R.; A-Moneim, M.T.

    1979-01-01

    A hydrodynamic-structural response analysis of the Mark-IICB loop was performed for the TREAT (Transient Reactor Test Facility) test AX-1. Test AX-1 is intended to provide information concerning the potential for a vapor explosion in an advanced-fueled LMFBR. The test will be conducted in TREAT with unirradiated uranium-carbide fuel pins in the Mark-IICB integral flowing sodium loop. Our analysis addressed the ability of the experimental hardware to maintain its containment integrity during the reference accident postulated for the test. Based on a thermal-hydraulics analysis and assumptions for fuel-coolant interaction in the test section, a pressure pulse of 144 MPa maximum pressure and pulse width of 1.32 ms has been calculated as the reference accident. The response of the test loop to the pressure transient was obtained with the ICEPEL and STRAW codes. Modelling of the test section was completed with STRAW and the remainder of the loop was modelled by ICEPEL

  13. Bacterial community analysis in upflow multilayer anaerobic reactor treating high-solids organic wastes.

    Science.gov (United States)

    Cho, Si-Kyung; Jung, Kyung-Won; Kim, Dong-Hoon; Kwon, Joong-Chun; Ijaz, Umer Zeeshan; Shin, Seung Gu

    2017-09-01

    A novel anaerobic digestion configuration, the upflow multi-layer anaerobic reactor (UMAR), was developed to treat high-solids organic wastes. The UMAR was hypothesized to form multi-layer along depth due to the upflow plug flow; use of a recirculation system and a rotating distributor and baffles aimed to assist treating high-solids influent. The chemical oxygen demand (COD) removal efficiency and methane (CH 4 ) production rate were 89% and 2.10 L CH 4 /L/d, respectively, at the peak influent COD concentration (110.4 g/L) and organic loading rate (7.5 g COD/L/d). The 454 pyrosequencing results clearly indicated heterogeneous distribution of bacterial communities at different vertical locations (upper, middle, and bottom) of the UMAR. Firmicutes was the dominant (>70%) phylum at the middle and bottom parts, while Deltaproteobacteria and Chloroflexi were only found in the upper part. Potential functions of the bacteria were discussed to speculate on their roles in the anaerobic performance of the UMAR system. © 2017 American Institute of Chemical Engineers Biotechnol. Prog., 33:1226-1234, 2017. © 2017 American Institute of Chemical Engineers.

  14. Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA

    International Nuclear Information System (INIS)

    William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-01-01

    The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10's initial criticality experiment for use as benchmarks for reactor physics codes

  15. Physics calculations for the Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Kalimullah; Kier, P.H.; Hummel, H.H.

    1977-06-01

    Calculations of distributions of power and sodium void reactivity, unvoided and voided Doppler coefficients and steel and fuel worths have been performed using diffusion theory and first-order perturbation theory for the LWR discharge Pu-fueled CRBR at BOL, the FFTF-grade Pu-fueled CRBR at BOL and for the beginning and end of equilibrium cycle of the LWR-Pu-fueled CRBR. The results of the burnup and breeding ratio calculations performed for obtaining the reactor compositions during the equilibrium cycle are also reported. Effects of sodium and steel contents on the distributions of sodium void reactivity and steel worth have also been studied. Errors and uncertainties in the reactivity coefficients due to cross-sections and the two-dimensional geometric representations of the reactor used in the calculations have also been estimated. Comparisons of the results with those in the CRBR PSAR are also discussed

  16. Physics-magnetics trade studies for tandem mirror reactors

    International Nuclear Information System (INIS)

    Campbell, R.B.; Perkins, L.J.; Blackfield, D.T.

    1985-01-01

    We describe and present results obtained from the optimization package of the Tandem Mirror Reactor Systems Code. We have found it to be very useful in searching through multidimensional parameter space, and have applied it here to study the effect of choke coil field strength and net electric power on cost of electricity (COE) and mass utilization factor (MUF) for MINIMARS type reactors. We have found that a broad optimum occurs at B/sub choke/ = 26 T for both COE and MUF. The COE economy of scale approaches saturation at quite low powers, around 600 MW(e). The saturation is mainly due to longer construction times for large plants, and the associated time related costs. The MUF economy of scale does not saturate, at least for powers up to 2400 MW(e)

  17. Nuclear Data Measurements for 21st Century Reactor Physics Applications

    International Nuclear Information System (INIS)

    Rahmat Aryaeinejad; Jerald D. Cole; Mark W. Drigert; James K. Jewell; Christopher A. McGrath; David W. Nigg; Edward L. Reber

    2003-01-01

    The United States Department of Energy (DOE), Office of Nuclear Energy (NE) has embarked on a long-term program to significantly advance the science and technology of nuclear energy. This is in response to the overall national plan for accelerated development of domestic energy resources on several fronts, punctuated by recent dramatic events that have emphasized the need for the US to reduce its dependence on foreign petroleum supplies. Key aspects of the DOE-NE agenda are embodied in the Generation-IV (Gen-IV) advanced nuclear energy systems development program and in the Advanced Fuel Cycle (AFC) program. The planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current nuclear power reactor systems as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The success of the overall NE effort will depend not only on sophisticated system development and engineering, but also on the advances in the supporting sciences and technologies. Of these, one of the most important is the improvement of the relevant fundamental nuclear science data bases, especially the evaluated neutron interaction cross section files that serve as the foundation of all reactor system designs, operating strategies, and fuel cycle engineering activities. The new concepts for reactors and fuel cycles involve the use of transuranic nuclides that were previously of little interest, and where experimentally measured information is lacking. The current state of the cross section database for some of these nuclides is such that design computations for advanced fast-spectrum reactor systems and fuel cycles that incorporate such materials in significant quantities are meaningful only for approximate conceptual applications. No actual system could reliably be designed according to currently accepted standards, nor

  18. Photocatalytic reactors for treating water pollution with solar illumination, Part 3: a simplified analysis for recirculating reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sagawe, G.; Bahnemann, D. [Hannover Univ. (Germany). Inst. fuer Technische Chemie; Brandi, R.J.; Cassano, A.E. [Universidad Nacional de Litoral, Santa Fe (Argentina). Inst. de Desarrollo Tecnologico para la Imdustria Quimica

    2004-11-01

    A solar photoreactor operated in the batch, recirculating mode is analyzed in terms of very simple observable variables such as the impinging photon flux, the incident area, the initial concentration, the flow rate, the reactor volume and a property defined as the Observed Photonic Efficiency. The proposed equipment is made of a tubular reactor, a tank, a pump and the connecting pipes. The analysis is formulated in terms of the photon input corresponding to an equivalent batch system that is derived as a new reaction coordinate for photoreactions. Employing several plausible approximations, the pollutant concentration evolution in the tank is cast in terms of very simple analytical solutions. Process photonic efficiencies are defined for the system operation and calculated with respect to the maximum achievable yield corresponding to the differential operation of the solar recirculating reactor. (Author)

  19. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  20. Multimedia Course on Nuclear Reactors Physics, Application to a Tailored On the Job Training Course

    International Nuclear Information System (INIS)

    Dies, Javier

    2014-01-01

    In order to improve education and training quality, a Multimedia on Nuclear Reactor Physics has been developed. In some institutions, this course is called Fundamentals of Nuclear Reactor Operation. Nowadays, this multimedia has about 800 slides and the text is in Spanish, English, French and Russian. Until now about 126 institutions from 53 countries have applied for the multimedia. The teacher uses the multimedia during his lectures. Students use it at home to study this course

  1. Proceedings of the 1992 topical meeting on advances in reactor physics

    International Nuclear Information System (INIS)

    1992-01-01

    This document, Volume 1, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Code Benchmarks and Validation; Fuel Management; Nodal Methods for Diffusion Theory; Criticality Safety and Applications and Waste; Core Computational Systems; Nuclear Data; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual papers have been cataloged separately. (FI)

  2. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  3. The startup performance and microbial distribution of an anaerobic baffled reactor (ABR) treating medium-strength synthetic industrial wastewater.

    Science.gov (United States)

    Jiang, Hao; Nie, Hong; Ding, Jiangtao; Stinner, Walter; Sun, Kaixuan; Zhou, Hongjun

    2018-01-02

    In this study, an anaerobic baffled reactor (ABR) with seven chambers was applied to treat medium-strength synthetic industrial wastewater (MSIW). The performance of startup and shock test on treating MSIW was investigated. During the acclimation process, the chemical oxygen demand (COD) of MSIW gradually increased from 0 to 2,000 mg L -1 , and the COD removal finally reached 90%. At shock test, the feeding COD concentration increased by one-fifth and the reactor adapted very well with a COD removal of 82%. In a stable state, Comamonas, Smithella, Syntrophomonas and Pseudomonas were the main populations of bacteria, while the predominant methanogen was Methanobacterium. The results of chemical and microbiological analysis indicated the significant advantages of ABR, including buffering shocks, separating stages with matching microorganisms and promoting syntrophism. Meanwhile, the strategies for acclimation and operation were of great importance. Further work can test reactor performance in the treatment of actual industrial wastewater.

  4. Performance evaluation of full scale UASB reactor in treating stillage wastewater

    Directory of Open Access Journals (Sweden)

    A.Mirsepasi , H. R. Honary , A. R. Mesdaghinia, A. H. Mahvi , H. Vahid , H. Karyab

    2006-04-01

    Full Text Available Upflow anaerobic sludge blanket (UASB reactors have been widely used for treatment of industrial wastewater. In this study two full-scale UASB reactors were investigated. Volume of each reactor was 420 m3. Conventional parameters such as pH, temperature and efficiency of COD, BOD, TOC removal in each reactor were investigated. Also several initial parameters in designing and operating of UASB reactors, such as upflow velocity, organic loading rate (OLR and hydraulic retention time were investigated. After modifying in operation conditions in UASB-2 reactor, average COD removal efficiency at OLR of 10–11 kg COD / m3 day was 55 percent. In order to prevent solids from settling, upflow velocity was increased to 0.35 m/h. Also to prevent solids from settling, the hydraulic retention time of wastewater in UASB-2 reactor was increased from 200 to 20 hours. This was expected that with good operation of UASB-2 reactor and with expanding of granules in the bed of the reactor, COD removal efficiency will be increased to more than 80 percent. But, because of deficiency on granulation and operation in UASB-2 reactor, this was not achieved. COD removal efficiency in the UASB-1 reactor was little. To enhance COD efficiency of UASB-1 reactor, several parameters were needed to be changed. These changes included enhancing of OLRs and upflow velocity, decreasing hydraulic retention time and operating with new sludge.

  5. Proceedings of the workshop on nuclear reaction data and nuclear reactors: Physics, design and safety

    International Nuclear Information System (INIS)

    Oblozinsky, P.; Gandini, A.

    1999-01-01

    The objective of the work shop organized by IAEA in cooperation with ICTP, Trieste and ENEA, Rome was to train scientists and engineers, particularly from developing countries, in modern reactor theory, nuclear data production and data use, with particular emphasis on applications in nuclear reactor physics, design and safety. This type of training is of special importance in the era of decreasing support to nuclear reactor activities in many countries, with an unfortunate consequence of vanishing infrastructure and expertise. In fact, the Workshop represents, worldwide, the only forum where scientists and engineers can get extensive and up-to-date information on nuclear reaction data, including physical background and evaluation methodology, and their application in nuclear reactor calculations. The proceedings is arranged in three parts according to the main topics of the Workshop. Part 1 deals with nuclear reactor models, including neutron resonances, fission optical model, statistical and preequilibrium models as well as nuclear level densities. Part 2 is devoted to nuclear data filing and processing, including nuclear data evaluation, and formatting, data libraries and services, and nuclear data processing codes. Part 3 is devoted to physics of nuclear reactors

  6. Reactor physics analysis of the pin-cell Doppler effect in a thermal nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kruijf, W.J.M. de

    1995-01-01

    This report has also been published as a PhD thesis. It deals with the Doppler effect in thermal nuclear reactors. Especially the behaviour of the reactor in transient conditions is an important issue. During such a transient the radial temperature profile in a fuel pin changes. In this PhD research effective fuel temperatures have been calculated for arbitrary temperature profiles in the fuel pin with the improved slowing-down code ROLAIDS-CPM. A general expression for the effective fuel temperature in a specific fuel pin is found by defining this effective fuel temperature as a weighted sum of the temperatures in different radial fuel zones. Also, the radial power profile in a fuel pin has been calculated by performing detailed burnup calculations, which agree very well with experimental data. (orig.).

  7. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  8. A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-06-01

    The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

  9. A stochastic physical-mathematical method for reactor kinetics analysis

    International Nuclear Information System (INIS)

    Velickovic, Lj.

    1966-01-01

    The developed theoretical model is concerned with BF 3 counter placed in the core of a low power reactor (a few MW) where statistical neutron effects are most evident. Our experiments were somewhat different. The detector used was and ionization chamber with double sampling, in ADC and in the time analyzer. The objective of this model was not to obtain precise numerical calculations, but to explain the method and the essentials of the correlation. Introducing all the six groups of delayed neutrons and possibly photoneutrons the model could be improved to obtained more realistic results

  10. Preliminary physics calculations for the Clinch River Breeder Reactor

    International Nuclear Information System (INIS)

    Kalimullah.

    1975-01-01

    Calculations of sodium void, fuel, and clad worths, power distribution, and control rod worths have been carried out for an R-Z model of the CRBR, using diffusion theory and first-order perturbation theory for material worths. The power distribution and control rod worths have also been calculated in two-dimensional triangular mesh geometry. The present results are preliminary because of inaccuracy of the reactor model and the cross sections used, but the final results are not expected to be greatly different. (U.S.)

  11. Proceedings of 2. Yugoslav symposium on reactor physics, Part 2, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant

  12. Review of PSI studies on reactor physics and thermal fluid dynamics of pebble bed reactors

    International Nuclear Information System (INIS)

    Prasser, Horst-Michael

    2014-01-01

    Switzerland is member of the Generation IV International Forum (GIF). The related work takes entirely place at PSI in the working groups of Gas-Cooled Fast Reactors and Very High Temperature Reactors. In the past, PSI has performed experimental and theoretical studies on criticality issues of pebble beds at the PROTEUS reactor, as well as a preliminary risk assessment of a prototypal HTR as an input for a comparison of energy supply options. PROTEUS was a critical assembly with an annular driver zone. The central region was filled by arrangements of fuel spheres. The reactivity effect of a water ingress was investigated by simulating the water by polyethylene rods of different diameter inserted into the gaps of a regular package. For sub-criticality measurements in pebble beds, a built-in pulsed neutron source was used. The experimental results were used to validate diffusion and higher order neutron transport models. Concerning thermal hydraulics of gas flows, the vast experience of PSI is focused on hydrogen transport, accumulation, and dispersion in containments of light water reactors. The phenomena are comparable in many aspects to the fluid dynamic issues relevant to HTR. Experiments on hydrogen flows are performed for numerous scenarios in the large-scale containment test facility PANDA. Hydrogen is substituted by helium as a model fluid. An important generic aspect is turbulent mixing in the presence of strong stratification, which is relevant for HTR as well. In a parallel project, generic small-scale mixing experiments with a high density ratio of 1:7 are carried out in a horizontal rectangular channel, where helium and nitrogen flows are brought into contact downstream of the rear edge of a splitter plate. Due to the high density ratio, turbulent mixing is affected by strong non-Boussinesq effects. The measurements taken by Particle Imaging Velocimetry (PIV) and Laser Induced Fluorescence techniques are compared to RANS and LES simulations. Similar large

  13. Revisiting basics of fertile-free-fuel reactor physics

    International Nuclear Information System (INIS)

    Fridman, Emil; Kolesnikov, Sergei; Galperin, Alex; Shwageraus, Eugene

    2006-01-01

    This paper investigates the basic feasibility of using Reactor Grade Pu in Fertile Free Fuel (FFF) matrix in Pressurized Water Reactors. Although the subject was addressed in the past, no systematic approach for assessing the main design challenges of FFF cores was published. In this work, we examine all commercially available burnable poison materials in all geometrical arrangements currently used by the nuclear industry with regards to their potential to alleviate the problems associated with the use of FFF in PWRs. The analysis was performed with a neutron transport and fuel assembly burnup code BOXER. Modified Linear Reactivity Model was applied to the two dimensional single fuel assembly results to approximate the full core characteristics. Based on the results of the performed analyses, Pu loaded FFF showed potential feasibility to be used in existing PWRs. All FFF problems may be significantly mitigated through the correct choice of BP material and configuration, while the final assessment on viability of full FFF core may be performed only on the basis of detailed 3D-full core analysis and the results of main design base transients. It was found that a combination of BP materials and geometries may be required to meet all FFF design goals. The use of enriched (in most effective isotope) burnable poisons, such as Er-167 and Gd- 157, may further improve the BP effectiveness and reduce the fuel cycle length penalty associated with their use. (authors)

  14. Reactor physics calculations for the control of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    Difilippo, F.C.; Abu-Shehadeh, M.; Perez, R.B.

    1988-01-01

    Efficient production of extremely high fluxes requires compact cores with consequent high power densities and initial excess reactivities. Strong space dependent neutron spectras and limited access to the small core are other characteristics that make design of the control system of these type of facilities an interesting problem. We present calculations of the worths of 10 B to reduce the initial excess reactivity, the worth of Hf and B control rods, and the neutron lifetimes, for the case of candidate designs for the Advanced Neutron Source reactor. 4 refs., 4 figs., 2 tabs

  15. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  16. Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

    Directory of Open Access Journals (Sweden)

    Giuseppe Palmiotti

    2012-01-01

    Full Text Available The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.

  17. Fast burner reactor benchmark results from the NEA working party on physics of plutonium recycle

    International Nuclear Information System (INIS)

    Hill, R.N.; Wade, D.C.; Palmiotti, G.

    1995-01-01

    As part of a program proposed by the OECD/NEA Working Party on Physics of Plutonium Recycling (WPPR) to evaluate different scenarios for the use of plutonium, fast reactor physics benchmarks were developed; fuel cycle scenarios using either PUREX/TRUEX (oxide fuel) or pyrometallurgical (metal fuel) separation technologies were specified. These benchmarks were designed to evaluate the nuclear performance and radiotoxicity impact of a transuranic-burning fast reactor system. International benchmark results are summarized in this paper; and key conclusions are highlighted

  18. Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Rubindamayugi, M.S.T.; Salakana, L.K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

    1997-12-31

    The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

  19. Students' assessment of interactive distance experimentation in nuclear reactor physics laboratory education

    Science.gov (United States)

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-10-01

    Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research reactors limit their accessibility to few educational programmes around the world. The concept of the Internet Reactor Laboratory (IRL) was introduced earlier as a new approach that utilises distance education in nuclear reactor physics laboratory education. This paper presents an initial assessment of the implementation of the IRL between the PULSTAR research reactor at North Carolina State University in the USA and the Department of Nuclear Engineering at Jordan University of Science and Technology (JUST) in Jordan. The IRL was implemented in teaching the Nuclear Reactor laboratory course for two semesters. Feedback from surveyed students verifies that the outcomes attained from using IRL in experimentation are comparable to that attainable from other on-campus laboratories performed by the students.

  20. Implications of nuclear physics in the development of Fast Breeder Reactors

    International Nuclear Information System (INIS)

    Rapeanu, S.; Ilie, P.; Vasiliu, G.; Popescu, C.; Boeriu, S.; Constantinescu, D.; Mateescu, S.

    1980-08-01

    The purpose of this paper is to point out the involved aspects of nuclear physics in the calculation and design of the fast reactors. After a brief description of the advantages of using the fast reactors in the national economy, the national programs concerning this activity are presented. The structure and operation conditions of the liquid metal fast breeder reactor (LMFBR) are also reviewed. Then, the methods aimed to calculate the core, the burn-up, the reactor dynamics, the analysis of accidents, the shielding, as well as, the materials required in the fast reactor calculation, are shortly given. Further on, it deals with the nuclear data types connected to the fast reactor calculations, with accuracy requirements for nuclear data, as well as, with the present stage of nuclear data for fissile, fertile and structural materials. The requirements for new differential data measurements, new integral data and benchmark experiments are presented. Data adjustement methods are also summarized. Some aspects of the structural material behaviour in intense gamma radiation and neutron fields existing into a fast reactor are also presented in the last part of this paper. The concluding remarks are mentioned at the end of the paper. (author)

  1. Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Obabko, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States); Tautges, Timothy [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, Robert Mark [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-21

    This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models of a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.

  2. Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

    2015-03-01

    In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.

  3. Membrane distillation combined with an anaerobic moving bed biofilm reactor for treating municipal wastewater.

    Science.gov (United States)

    Kim, Hyun-Chul; Shin, Jaewon; Won, Seyeon; Lee, Jung-Yeol; Maeng, Sung Kyu; Song, Kyung Guen

    2015-03-15

    A fermentative strategy with an anaerobic moving bed biofilm reactor (AMBBR) was used for the treatment of domestic wastewater. The feasibility of using a membrane separation technique for post-treatment of anaerobic bio-effluent was evaluated with emphasis on employing a membrane distillation (MD). Three different hydrophobic 0.2 μm membranes made of polytetrafluoroethylene (PTFE), polyvinylidene fluoride (PVDF), and polypropylene (PP) were examined in this study. The initial permeate flux of the membranes ranged from 2.5 to 6.3 L m(-2) h(-1) when treating AMBBR effluent at a temperature difference between the feed and permeate streams of 20 °C, with the permeate flux increasing in the order PP membrane gradually decreased to 84% of the initial flux after the 45 h run for distillation, while a flux decline in MD with either the PVDF or PP membrane was not found under the identical distillation conditions. During long-term distillation with the PVDF membrane, total phosphorus was completely rejected and >98% rejection of dissolved organic carbon was also achieved. The characterization of wastewater effluent organic matter (EfOM) using an innovative suite of analytical tools verified that almost all of the EfOM was rejected via the PVDF MD treatment. Copyright © 2015 Elsevier Ltd. All rights reserved.

  4. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  5. Reactor physics ideas to design novel reactors with faster fissile growth

    International Nuclear Information System (INIS)

    Jagannathan, V.; Pal, U.; Karthikeyan, R.; Raj, D.; Srivastava, A.; Khan, S. A.

    2007-01-01

    There are several types of fission reactors operating in the world adopting generally the open fuel cycle which considers the naturally available fissile nuclide, viz., 2 35U. The accumulated discharged fuel is considered as waste in some countries. However the discharged fuel contains the precious man-made fissile plutonium which would provide the sole means of harnessing the nuclear energy from either depleted uranium or the natural thorium in future. It must be emphasized that the present day power reactors use just about 0.5% of the mined uranium and it would be imprudent to discard the rest of the mass as waste. It is therefore necessary to explore ways and means of exploiting the fertile mass which has the potential of providing the energy without the green house effects for millennia to come. This has to be done by innovating means of large scale fertile to fissile conversion and then using the man-made fissile material for sustenance as well as growth of fission nuclear power. This paper attempts to give a broad picture of the available options and the challenges in realizing the theoretical possibilities

  6. Microbial dynamics in anaerobic digestion reactors for treating organic urban residues during the start-up process.

    Science.gov (United States)

    Alcántara-Hernández, R J; Taş, N; Carlos-Pinedo, S; Durán-Moreno, A; Falcón, L I

    2017-06-01

    , Firmicutes and Spirochaetes. These results clarify the bacterial processes behind new reactors establishment for treating organic wastes in urban areas. © 2017 The Society for Applied Microbiology.

  7. Identification of the physical parameters of a nuclear reactor core by a dynamic method

    International Nuclear Information System (INIS)

    Cervoni, C.

    1981-10-01

    The aim of this thesis was to qualify an identification and measuring method of the physical parameters of a nuclear reactor core, that is to say the integral antireactivity of control rods, as also the temperature coefficients and the thermal exchange coefficient. This method can be applied to PWR type reactors. The method used for the present study is the power track measuring method. It is applied (1) to measure the anti-reactivity of the (N-1) control assemblies of a PWR reactor, (2) to measure the antireactivity of the control rods in the rod drop case on the experimental reactor ''ORPHEE'', (3) to the identification of the thermal feedback coefficients in the rod drop case. One shows how the detectors have to be used to solve the spatial effect problem for these different experiments [fr

  8. Investigation for calculation methods used in analyzing the physics characteristics of nuclear power reactor

    International Nuclear Information System (INIS)

    Nguyen Tuan Khai; Hoang Van Khanh; Phan Quoc Vuong; Tran Viet Phu; Tran Vinh Thanh; Nguyen Thi Mai Huong; Nguyen Thi Dung; Le Tran Chung; Nguyen Minh Tuan; Tran Quoc Duong

    2014-01-01

    The project aims at nuclear human resource development and enhancement in research capability in reactor physics and kinetics at Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat). The main research items of the project can be summarized as follows: i) Considering possibility on using modern calculation techniques and methods in investigating neutronic characteristics and neutronics-thermal hydraulics coupling. This item is proposed to carry out based on international collaboration with Prof. Le Trong Thuy, San Jose University, US; ii) Carrying out the collaborative activities in research and training between Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat); iii) Opening two-week training course on nuclear reactor engineering (25 Nov - 12 Dec 2013) in collaboration with Japan Atomic Energy Agency (JAEA). (author)

  9. Physics considerations of the Reversed-Field Pinch fusion reactor

    International Nuclear Information System (INIS)

    Hagenson, R.L.; Krakowski, R.A.

    1980-01-01

    A conceptual engineering design of a fusion reactor based on plasma confinement in a toroidal Reversed-Field Pinch (RFP) configuration is described. The plasma is ohmically ignited by toroidal plasma currents which also inherently provide the confining magnetic fields in a toroidal chamber having major and minor radii of 12.7 and 1.5 m, respectively. The DT plasma ignites in 2 to 3 s and undergoes a transient, unrefueled burn at 10 to 20 keV for approx. 20 s to give a DT burnup of approx. 50%. Accounting for all major energy sinks yields a cost-optimized system with a recirculating power fraction of 0.17; the power output is 750 MWe

  10. Reactor physics studies in the steam flooded GCFR-Phase II critical assembly

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.

    1978-08-01

    A possible accident scenario in a Gas-Cooled Fast Reactor (GCFR) is the leakage of secondary steam into the core. Considerable analytical effort has gone into the study of the effects of such an accidental steam entry. The work described represents the first full scale experimental study of the steam-entry phenomenon in GCFRs. The reference GCFR model used for the study was the benchmark GCFR Phase II assembly, and polyethylene foam was used to provide a very homogeneous steam simulation. The reactivity worth of steam entry was measured for three different steam densities. In addition, a set of integral physics parameters were measured in the largest steam density (0.008 g/cm 3 ) configuration. The corresponding parameters were also measured in dry reference GCFR critical assembly for comparison. The experiments were analyzed using ENDF/B-IV data and two-dimensional diffusion theory methods. As in earlier GCFR critical experiments analysis, the Benoist method was used to treat the problem of neutron streaming

  11. A simplified analysis of granule behavior in ASBR and UASB reactors treating low-strength synthetic wastewater

    Directory of Open Access Journals (Sweden)

    R. G. Veronez

    2005-09-01

    Full Text Available This work presents an analysis of the changes observed in granule characteristics of sludge in the treatment of synthetic wastewater at a concentration of about 500 mgCOD/L in batch, fed-batch (ASBR and continuous (UASB bench-scale reactors under similar experimental conditions. Physical and microbiological properties of the granules were characterized as average particle size and sedimentation time and by optical and epifluorescence microscopy. Several samples were analyzed in order to identify the morphologies. Granules from sequencing batch and fed-batch reactors, either with or without mechanical mixing, did not undergo any physical or microbiological changes. However, during the experiment granules from the UASB reactor agglomerated due to the formation and accumulation of a viscous material, probably of microbial origin, when operated at low superficial velocities (0.072, 0.10 and 0.19 m/h. When the superficial velocity was increased to 8.0-10.0 m/h by means of liquid-phase recirculation, the granules from the UASB reactor underwent flocculation and the microbiological characteristics changed in such a way that the equilibrium of microbial diversity in the inoculum was not maintained. As a result, the only reactor that maintained efficiency and good solids retention during the assays was the ASBR, showing that there is a correlation between maintenance of microbial diversity and operating mode in the case of anaerobic treatment of low-strength wastewaters.

  12. Production and characterization of scum and its role in odour control in UASB reactors treating domestic wastewater.

    Science.gov (United States)

    Souza, C L; Silva, S Q; Aquino, S F; Chernicharo, C A L

    2006-01-01

    There are few studies in the literature that have aimed at characterizing the physical, chemical, and microbial aspects of scum produced in UASB reactors. In addition, there is little information on the influence of operational conditions of UASB reactors on scum formation, and the present work addresses these issues. Three demo-scale UASB reactors, fed on domestic wastewater, were employed to monitor the formation and its characteristics. Scum production was periodically assessed during different operational phases, and its characterization involved analyses of BOD, COD, solids, sulfide, sulfate, microscopic observations, as well as biodegradability tests. The results show that the scum formed was physically, chemically, and microscopically similar in both geminated reactors, being comprised mainly of organic material of low biodegradability. Several bacterial morphotypes, mainly filaments and rods, with internal sulfur granules, were observed, and the aerobic microorganisms that developed at the scum layer as a result of photosynthetic activity of cyanobacteria, seemed to play an important role in sulfide removal and odour control. Scum production rates were similar in both reactors, but the imposed higher upflow velocities resulted in a higher production rate and in a reduced biodegradability of the scum.

  13. Methodology for development of health physics procedures at research reactors in agreement states

    International Nuclear Information System (INIS)

    Woodard, R.C.; Bauer, T.L.; Wehring, B.W.

    1991-01-01

    The University of Texas at Austin is awaiting final license approval to operate a new 1 MW TRIGA reactor for teaching and research. All reactor and laboratory operations, experiments, and monitoring are carried out under health physics procedures that address to ensure consideration of all applicable documents as references in order to comply with the regulations and accepted good practices. This paper examines the development of one procedure Radioactive Material Control by use of the method. The process is examined as a tool to apply to any health physics procedure development. Further discussion focuses on the regulatory anomalies observed during development of the procedure and presents the arguments for the authors resolution of these issues. The design of the reactor facility is also detailed to allow for understanding of the problems encountered during procedural development

  14. Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    1991-01-01

    Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)

  15. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  16. Kinetic modelling and characterization of microbial community present in a full-scale UASB reactor treating brewery effluent.

    Science.gov (United States)

    Enitan, Abimbola M; Kumari, Sheena; Swalaha, Feroz M; Adeyemo, J; Ramdhani, Nishani; Bux, Faizal

    2014-02-01

    The performance of a full-scale upflow anaerobic sludge blanket (UASB) reactor treating brewery wastewater was investigated by microbial analysis and kinetic modelling. The microbial community present in the granular sludge was detected using fluorescent in situ hybridization (FISH) and further confirmed using polymerase chain reaction. A group of 16S rRNA based fluorescent probes and primers targeting Archaea and Eubacteria were selected for microbial analysis. FISH results indicated the presence and dominance of a significant amount of Eubacteria and diverse group of methanogenic Archaea belonging to the order Methanococcales, Methanobacteriales, and Methanomicrobiales within in the UASB reactor. The influent brewery wastewater had a relatively high amount of volatile fatty acids chemical oxygen demand (COD), 2005 mg/l and the final COD concentration of the reactor was 457 mg/l. The biogas analysis showed 60-69% of methane, confirming the presence and activities of methanogens within the reactor. Biokinetics of the degradable organic substrate present in the brewery wastewater was further explored using Stover and Kincannon kinetic model, with the aim of predicting the final effluent quality. The maximum utilization rate constant U max and the saturation constant (K(B)) in the model were estimated as 18.51 and 13.64 g/l/day, respectively. The model showed an excellent fit between the predicted and the observed effluent COD concentrations. Applicability of this model to predict the effluent quality of the UASB reactor treating brewery wastewater was evident from the regression analysis (R(2) = 0.957) which could be used for optimizing the reactor performance.

  17. Treating municipal solid waste leachate in a pilot scale upflow anaerobic sludge blanket reactor under tropical temperature

    Directory of Open Access Journals (Sweden)

    Abbas Alizadeh Shooshtari

    2012-01-01

    Full Text Available Aims: The objective of this study was to investigate an Upflow Anaerobic Sludge Blanket (UASB reactor efficiency in treating municipal landfill leachate, under tropical temperature. Materials and Methods: A 30-liter pilot-scale UASB reactor was used to treat the municipal solid waste leachate, under tropical temperature, for 230 days. The reactor was inoculated with 10 liters of anaerobic sludge from an anaerobic digester, in an agro industry′s wastewater treatment plant. The Volatile Suspended Solids (VSS of sludge were 65 g/L, with volatile suspended solids to suspended solids (VSS/SS ratio of 0.74. The reactor was operated in mesophilic (34 - 39°C temperature. Results: After reaching a stable operation, the reactor was exposed to raw leachate, with mean chemical oxygen demand (COD concentrations of 35 g/L. The leachate was diluted to 9 - 10 g/L at Organic Loading Rates (OLRs of 2, 6, 12, 15 g COD/L.d and decreased again to 12 g COD/L.d, resulting in 45, 76, 84, 68, and 79% removal efficiency and increased again to 87% removal efficiency for COD, at Hydraulic Retention Times (HRTs of 6, 1.6, 0.83, and 0.67 days, respectively, in the UASB. In the reactor used in this study, the heavy metals were removed by adsorption on biomass, and the maximum removal rate was 68% for Zinc (Zn. Conclusions: It was concluded that the optimum OLR for diluted leachate up to 10 g COD/l, was 12 g COD/L.d at an HRT of 0.67 day (16 hours.

  18. Standard interface files and procedures for reactor physics codes. Version IV

    International Nuclear Information System (INIS)

    O'Dell, R.D.

    1977-09-01

    Standards, procedures, and recommendations of the Committee on Computer Code Coordination for promoting the exchange of reactor physics codes are updated to Version IV status. Standards and procedures covering general programming, program structure, standard interface files, and file management and handling subroutines are included

  19. Engineering and physics considerations for a linear theta-pinch hybrid reactor (LTPHR)

    International Nuclear Information System (INIS)

    Krakowski, R.A.; Miller, R.L.; Hagenson, R.L.

    1976-01-01

    A fusion-fission hybrid reactor based on pulsed, high-β, linear theta-pinch magnetic confinement is considered. A preliminary design which incorporates key physics, engineering and economic considerations is presented. An extensive presentation of the system energy balance is made, and this energy balance is evaluated parametrically. The feasibility of end-loss reduction is addressed

  20. 75 FR 62695 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2010-10-13

    ... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule. SUMMARY: The... nuclear fuel in transit? H. Why require a telemetric position monitoring system or an alternative tracking... nuclear fuel in transit. The interim final rule added 10 CFR 73.37, ``Requirements for Physical Protection...

  1. Neutron physics and nuclear data measurements with accelerators and research reactors

    International Nuclear Information System (INIS)

    1988-08-01

    The report contains a collection of lectures devoted to the latest theoretical and experimental developments in the field of fast neutron physics and nuclear data measurements. The possibilities offered by particle accelerators and research reactors for research and technological applications in these fields are pointed out. Refs, figs and tabs

  2. The integral fast reactor (IFR) concept: Physics of operation and safety

    International Nuclear Information System (INIS)

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

  3. Development of intelligent physical start-up system for nuclear reactor

    International Nuclear Information System (INIS)

    Wang Canhui; Li Xiang; Huang Liyuan; Fu Guoen; Hu Hai

    2008-01-01

    In this paper, the Intelligent physical start-up system for nuclear reactor introduced the system composing, hardware design and software design. The system has some merits such as handy operation, fast and accurate mathematic and nicer human-machine interface. (authors)

  4. Physical inventory verification exercise at a light-water reactor facility

    International Nuclear Information System (INIS)

    Bosler, G.E.; Menlove, H.O.; Halbig, J.K.

    1986-04-01

    A simulated physical inventory verification exercise was performed at the Three Mile Island (TMI) Unit 1 reactor. Inspectors from the Internatinal Atomic Energy Agency made measurements on fresh- and spent-fuel assemblies and verified the special nuclear material inventory at TMI. Simulated inspection log sheets and computerized inspection reports were prepared

  5. The integral fast reactor (IFR) concept: physics of operation and safety

    International Nuclear Information System (INIS)

    Wade, D.C.; Chang, Y.I.

    1987-01-01

    The IFR concept employs a pool layout, a U/Pu/Zr metal alloy fuel and a closed fuel cycle based on pyrometallurgical reprocessing and injection casting refabrication. The reactor physics issues of designing for inherent safety and for a closed fissile self-sufficient integral fuel cycle with uranium startup and potential actinide transmutation are discussed

  6. Application of linear and higher perturbation theory in reactor physics

    International Nuclear Information System (INIS)

    Woerner, D.

    1978-01-01

    For small perturbations in the material composition of a reactor according to the first approximation of perturbation theory the eigenvalue perturbation is proportional to the perturbation of the system. This assumption is true for the neutron flux not influenced by the perturbance. The two-dimensional code LINESTO developed for such problems in this paper on the basis of diffusion theory determines the relative change of the multiplication constant. For perturbations varying the neutron flux in the space of energy and position the eigenvalue perturbation is also influenced by this changed neutron flux. In such cases linear perturbation theory yields larger errors. Starting from the methods of calculus of variations there is additionally developed in this paper a perturbation method of calculation permitting in a quick and simple manner to assess the influence of flux perturbation on the eigenvalue perturbation. While the source of perturbations is evaluated in isotropic approximation of diffusion theory the associated inhomogeneous equation may be used to determine the flux perturbation by means of diffusion or transport theory. Possibilities of application and limitations of this method are studied in further systematic investigations on local perturbations. It is shown that with the integrated code system developed in this paper a number of local perturbations may be checked requiring little computing time. With it flux perturbations in first approximation and perturbations of the multiplication constant in second approximation can be evaluated. (orig./RW) [de

  7. Neutron standard cross sections in reactor physics - Need and status

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1990-01-01

    The design and improvement of nuclear reactors require detailed neutronics calculations. These calculations depend on comprehensive libraries of evaluated nuclear cross sections. Most of the cross sections that form the data base for these evaluations have been measured relative to neutron cross-section standards. The use of these standards can often simplify the measurement process by eliminating the need for a direct measurement of the neutron fluence. The standards are not known perfectly, however; thus the accuracy of a cross-section measurement is limited by the uncertainty in the standard cross section relative to which it is measured. Improvements in a standard cause all cross sections measured relative to that standard to be improved. This is the reason for the emphasis on improving the neutron cross-section standards. The continual process of measurement and evaluation has led to improvements in the accuracy and range of applicability of the standards. Though these improvements have been substantial, this process must continue in order to obtain the high-quality standards needed by the user community

  8. Reactor physics tests and benchmark analyses of STACY

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Umano, Takuya

    1996-01-01

    The Static Experiment Critical Facility, STACY in the Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF is a solution type critical facility to accumulate fundamental criticality data on uranyl nitrate solution, plutonium nitrate solution and their mixture. A series of critical experiments have been performed for 10 wt% enriched uranyl nitrate solution using a cylindrical core tank. In these experiments, systematic data of the critical height, differential reactivity of the fuel solution, kinetic parameter and reactor power were measured with changing the uranium concentration of the fuel solution from 313 gU/l to 225 gU/l. Critical data through the first series of experiments for the basic core are reported in this paper for evaluating the accuracy of the criticality safety calculation codes. Benchmark calculations of the neutron multiplication factor k eff for the critical condition were made using a neutron transport code TWOTRAN in the SRAC system and a continuous energy Monte Carlo code MCNP 4A with a Japanese evaluated nuclear data library, JENDL 3.2. (J.P.N.)

  9. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    This includes computer codes for burnup studies based on transport theory and Monte. Carlo methods, codes for studying the kinetics of ADS and sub-critical facilities driven by 14 MeV neutron generators for ADS experiments and development of sub-criticality measurement methods. The paper discusses the physics ...

  10. Start-up of an anaerobic hybrid (UASB/filter) reactor treating wastewater from a coffee processing plant.

    Science.gov (United States)

    Bello-Mendoza, R; Castillo-Rivera, M F

    1998-10-01

    The ability of an anaerobic hybrid reactor, treating coffee wastewater, to achieve a quick start-up was tested at pilot scale. The unacclimatized seed sludge used showed a low specific methanogenic activity of 26.47 g CH4 as chemical oxygen demand (COD)/kg volatile suspended solids (VSS) x day. This strongly limited the reactor performance. After a few days of operation, a COD removal of 77.2% was obtained at an organic loading rate (OLR) of 1.89 kg COD/m3 x day and a hydraulic retention time (HRT) of 22 h. However, suddenly increasing OLR above 2.4 kg COD/m3 x day resulted in a deterioration in treatment efficiency. The reactor recovered from shock loads after shutdowns of 1 week. The hybrid design of the anaerobic reactor prevented the biomass from washing-out but gas clogging in the packing material was also observed. Wide variations in wastewater strength and flow rates prevented stable reactor operation in the short period of the study.

  11. Benchmark of physics design of a proposed 30 MW Multi Purpose Research Reactor using a Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Singh, Tej; Kumar, Jainendra; Sharma, Archana; Singh, Kanchhi; Raina, V.K.; Srinivasan, P.

    2009-01-01

    At present Dhruva and Cirus reactors provide majority of research reactor based experimental/irradiation facilities to cater to various needs of the vast pool of researchers in the field of sciences research and development work for nuclear power plants and production of radioisotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 30 MWt Multi Purpose Research Reactor is proposed to be constructed. This paper describes some of the physics design features of this reactor using MCNP code to validate the deterministic methods. The criticality calculations for 100 material testing reactor (JHR) of France and 610 MW SAVANNAH thermal reactor were performed using MCNP computer codes to boost the confidence level in designing the physics design of reactor core. (author)

  12. Guidance of operation practice and reactor physics experiments using JRR-4

    International Nuclear Information System (INIS)

    Yokoo, Kenji; Horiguchi, Hironori; Yagi, Masahiro; Nagadomi, Hideki; Yamamoto, Kazuyoshi; Sasajima, Fumio; Ohyama, Koji; Ishikuro, Yasuhiro; Sasaki, Tsutomu; Hirane, Nobuhiko; Kimura, Kazuya; Arai, Nobuyoshi

    2007-03-01

    Reactor operation training using JRR-4 (Japan Research Reactor No.4) was started in FY1969, as one of the curriculums of Nuclear Technology and Education Center (NuTEC). Since then, the program was updated and carried out for reactor operation training, control rod calibration, and measurement of various kind of characteristics. JRR-4 has been contributed for nuclear engineer training for over 1,700 trainees from bother domestic and foreign countries. JRR-4 can be used for experiment from zero power to 3,500 kW, and the trainees can gain experience to operate the reactor from start up to shutdown, for not only zero-power experiments (critical approach, control rod calibration, reactivity measurement, etc.) but also other experiments under high power operation (xenon effect, temperature effects, reactor power calibration, etc.). This report is prepared as a standard text for training operation and experiments for reactor physics, based on various kinds of guidance texts being used for training purpose. (author)

  13. Treated LAW Feed Evaporation: Physical and Solubility Determination (U)

    International Nuclear Information System (INIS)

    Josephs, JE.

    2003-01-01

    Evaporation is employed in several places in the Waste Treatment Plant pretreatment process to minimize the volume of waste that must be treated in down-stream vitrification processes. Evaporation is the first unit process in pretreatment (Waste Feed Evaporators), applied before LAW vitrification (Treated Feed Evaporator), and concentrates ion exchange eluate (Cs Eluate Evaporator) prior to HLW vitrification. The goal of the Treated Feed Evaporation process is removal of the maximum water content without producing additional insoluble solids. Prior testing of evaporation systems for process feed was completed to support compliance with regulatory permits and to prepare a model of the evaporation system. These tests also indicated a marked tendency for foaming in the WTP evaporators. To date, evaporation testing and modeling have focused on the Treated feed and Cs eluate evaporation systems. This has been the first work performed that investigates evaporation of secondary-waste recycle streams in the Treated LAW Feed Evaporator. Secondary-waste recycles from the LAW off- gas scrubbing system have been the major contributors to the overall Treated Feed Evaporator recycle volume. Experience from Savannah River Site operations suggests that the introduction of silica- laden recycles to an evaporator along with high-sodium treated LAW can significantly increase the likelihood of forming sodium-alumina-silicate precipitates upon concentration. Furthermore, there is considerable interest on the part of the WTP project to evaluate the potential of projected evaporator feed and concentrate blends to produce sodium aluminosilicate precipitates

  14. Physical Therapy to Treat Torn Meniscus Comparable to Surgery for Many Patients

    Science.gov (United States)

    ... Spotlight on Research Physical Therapy to Treat Torn Meniscus Comparable to Surgery for Many Patients By Colleen ... involves surgically removing the torn part of the meniscus and stabilizing it, or physical therapy. However, it ...

  15. Treating domestic sewage by Integrated Inclined-Plate-Membrane bio-reactor

    Science.gov (United States)

    Song, Li Ming; Wang, Zi; Chen, Lei; Zhong, Min; Dong, Zhan Feng

    2017-12-01

    Membrane fouling shorten the service life of the membrane and increases aeration rate for membrane surface cleaning. Two membrane bio-reactors, one for working and another for comparing, were set up to evaluate the feasibility of alleviating membrane fouling and improving wastewater treatment efficiency by integrating inclined-plate precipitation and membrane separation. The result show that: (1) Inclined-plate in reactor had a good effect on pollutant removal of membrane bioreactor. The main role of inclined-plate is dividing reactor space and accelerating precipitation. (2) Working reactor have better performance in COD, TN and TP removal, which can attribute to that working reactor (integrated inclined-plate-Membrane bioreactor) takes both advantages of membrane separation and biological treatment. When influent COD, TP and TN concentration is 163-248 mg/L, 2.08-2.81 mg/L and 24.38-30.49 mg/L in working reactor, effluent concentration is 27-35 mg/L, 0.53-0.59 mg/L and 11.28-11.56 mg/L, respectively. (3) Membrane fouling was well alleviated in integrated inclined-plate-Membrane bioreactor, and membrane normal service time is significantly longer than that in comparing reactor, which can attribute to accelerating precipitation of inclined-plate. In summary, integrated inclined-plate-Membrane bioreactor is a promising technology to alleviating membrane fouling and improving wastewater treatment efficiency, having good performance and bright future in application.

  16. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

    International Nuclear Information System (INIS)

    2017-01-01

    The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

  17. Evaluation of a hybrid anaerobic biofilm reactor treating winery effluents and using grape stalks as biofilm carrier

    OpenAIRE

    Wahab, Mohamed Ali; Habouzit, Frédéric; Bernet, Nicolas; Jedidi, Naceur; Escudié, Renaud

    2016-01-01

    Wine production processes generate large amount of both winery wastewater and solid wastes. Furthermore, working periods, volumes and pollution loads greatly vary over the year. Therefore, it is recommended to develop a low cost treatment technology for the treatment of winery effluents taking into account the variation of the organic loading rate (OLR). Accordingly, we have investigated the sequential operation of an anaerobic biofilm reactor treating winery effluents and using grape stalks ...

  18. Review of the status of reactor physics predictive methods for burnable poisons in CAGRs

    International Nuclear Information System (INIS)

    Edens, D.J.; McEllin, M.

    1983-01-01

    An essential component of the design of Commercial Advanced Gas Cooled Reactor fuel necessary to achieve higher discharge irradiations is the incorporation of burnable poisons. The poisons enable the more highly enriched fuel required to reach higher irradiation to be loaded without increasing the peak channel power. The optimum choice of fuel enrichment and poison loading will be made using reactor physics predictive methods developed by Berkeley Nuclear Laboratories. The paper describes these methods and the evidence available to support them from theoretical comparisons, zero energy experiments, WAGR irradiations, and measurements on operating CAGR's. (author)

  19. General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics

    International Nuclear Information System (INIS)

    2001-10-01

    The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

  20. Physics considerations in the design of liquid metal reactors for transuranium element consumption

    International Nuclear Information System (INIS)

    Khalil, H.; Hill, R.; Fujita, E.; Wade, D.

    1992-01-01

    The management of transuranic nuclides in liquid metal reactors (LMR's) is considered based on the use of the Integral Fast Reactor (IFR) concept. Unique features of the IFR fuel cycle with respect to transuranic management are identified. These features are exploited together with the hard spectrum of LMR's to demonstrate the neutronic feasibility of a wide range of transuranic management options ranging from efficient breeding to pure consumption. Core physics aspects of the development of a low sodium void worth transuranic burner concept are described. Neutronics performance parameters and reactivity feedback characteristics estimated for this core concept are presented

  1. Coupling digestion in a pilot-scale UASB reactor and electrochemical oxidation over BDD anode to treat diluted cheese whey.

    Science.gov (United States)

    Katsoni, Alphathanasia; Mantzavinos, Dionissios; Diamadopoulos, Evan

    2014-11-01

    The efficiency of the anaerobic treatment of cheese whey (CW) at mesophilic conditions was investigated. In addition, the applicability of electrochemical oxidation as an advanced post-treatment for the complete removal of chemical oxygen demand (COD) from the anaerobically treated cheese whey was evaluated. The diluted cheese whey, having a pH of 6.5 and a total COD of 6 g/L, was first treated in a 600-L, pilot-scale up-flow anaerobic sludge blanket (UASB) reactor. The UASB process, which was operated for 87 days at mesophilic conditions (32 ± 2 °C) at a hydraulic retention time (HRT) of 3 days, led to a COD removal efficiency between 66 and 97 %, while the particulate matter of the wastewater was effectively removed by entrapment in the sludge blanket of the reactor. When the anaerobic reactor effluent was post-treated over a boron-doped diamond (BDD) anode at 9 and 18 A and in the presence of NaCl as the supporting electrolyte, complete removal of COD was attained after 3-4 h of reaction. During electrochemical experiments, three groups of organochlorinated compounds, namely trihalomethanes (THMs), haloacetonitriles (HANs), and haloketons (HKs), as well as 1,2-dichloroethane (DCA) and chloropicrin were identified as by-products of the process; these, alongside free chlorine, are thought to increase the matrix ecotoxicity to Artemia salina.

  2. ReactorHealth Physics operations at the NIST center for neutron research.

    Science.gov (United States)

    Johnston, Thomas P

    2015-02-01

    Performing health physics and radiation safety functions under a special nuclear material license and a research and test reactor license at a major government research and development laboratory encompasses many elements not encountered by industrial, general, or broad scope licenses. This article reviews elements of the health physics and radiation safety program at the NIST Center for Neutron Research, including the early history and discovery of the neutron, applications of neutron research, reactor overview, safety and security of radiation sources and radioactive material, and general health physics procedures. These comprise precautions and control of tritium, training program, neutron beam sample processing, laboratory audits, inventory and leak tests, meter calibration, repair and evaluation, radioactive waste management, and emergency response. In addition, the radiation monitoring systems will be reviewed including confinement building monitoring, ventilation filter radiation monitors, secondary coolant monitors, gaseous fission product monitors, gas monitors, ventilation tritium monitor, and the plant effluent monitor systems.

  3. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  4. Monitoramento de parâmetros físicos, químicos e biológicos em um reator anaeróbio híbrido (RAH em escala piloto, tratando água residuária do café produzido por via úmida Monitoring of physical, chemical, and biological parameters of a hibrid anaerobic reactor (HAR in pilot scale, treating wastewater from wet coffee production

    Directory of Open Access Journals (Sweden)

    Vivian Galdino da Silva

    2010-02-01

    Full Text Available Conduziu-se este trabalho, com o objetivo de fornecer informações a respeito da investigação experimental realizada na estação piloto no Núcleo de Estudos em Cafeicultura (NECAF/UFLA, para tratamento de água residuária do café (ARC especificamente, o monitoramento de um Reator Anaeróbio Híbrido (RAH, utilizando minifiltros preenchidos com dois tipos de meio suporte (argila expandida e seixo rolado. O RAH possuía fundo falso e foi utilizado como inóculo lodo anaeróbio de Upflow anaerobic sludge blanket reator (reator anaeróbio de manta de lodo - UASB (escala-plena, de esgoto doméstico. O sistema foi monitorado por meio de parâmetros operacionais e de análises químicas e físico-químicas. Foi constatado que a fase experimental em que se deu o experimento dificultou a eficiência do RAH por ter sido submetido a cargas inferiores às previstas, uma vez que o efluente passava por tratamento em outras unidades, minimizando assim, a concentração orgânica a ser tratada. O RAH apresentou concentrações afluentes médias de: 484; 168 e 92 mg DQOtot L-1 e concentrações efluentes médias de 344; 159 e 90 mg DQOtot L-1, para os tempos de detenção hidráulica (TDH de 28,5; 23,7 e 18,0 horas, respectivamente. O RAH apresentou equilíbrio com relação aos parâmetros medidos e boa estabilidade. O efluente analisado não apresentou riscos de salinidade, uma vez que a CE ficou na faixa, entre 0,70 e 3,0 dS m-1 e os SDT, entre 450 a 2000 mg L-1. Quanto à sodificação, a Razão de Adsorção de Sódio (RAS ficou entre 0 e 3 e a CE>0,7 dS m-1, podendo ser reutilizado para irrigação da maioria das culturas e solos.The purpose of this work is to supply information regarding to the experimental investigation carried out in a pilot scale system, located at Coffee Study Research Center (NECAF/UFLA. The research focused mainly on a Hybrid Anaerobic Reactor (HAR operating with mini-filters filled with two types of physical support

  5. Qinshan CANDU project simulation of reactor physics tests at low power

    International Nuclear Information System (INIS)

    Banica, C.; Tin, E.S.Y.; Mingjun, C.; Shad, M.A.; Schwanke, P.; Jenkins, D.A.

    2003-01-01

    Two new CANDU 6 reactors located in Qinshan, China, have recently been added to AECL's CANDU family. As a result of a very successful project, the first unit entered commercial operation in December 2002. As with all CANDU reactors, a series of physics tests were performed after first criticality was achieved. These tests were presimulated with the RFSP code and the results were compared to the measured data. The Phase-B commissioning is described in this paper, with an emphasis on lessons learned and quality of the fit of the measurements to the presimulations. The measured device reactivity worths in terms of changes in zone controller fills compared well with the results of the presimulations. Good agreement was also obtained between precalculated fluxes at detector locations and measured detector readings for all rundown tests. These results give confidence that the shutdown systems and reactor regulating system are functioning as expected and also provide validation of the Qinshan RFSP model. (author)

  6. Fast neutron breeder reactor Rapsodie - situation of physics, hydraulic, thermal and dynamics studies and studies of stability early in 1963

    International Nuclear Information System (INIS)

    1964-01-01

    Early in 1963, it was necessary to make a choice among the two fuels examined for Rapsodie: the UPuMo alloy with double cladding, Nb and stainless steel, and the UO 2 -PuO 2 mix oxide. This report presents the results of the studies effected with the two types of fuel. We reconsider at first the different models which have been studied and we give a detailed description of the alloy and oxide cores as they are envisaged early in 1963. We give then the most important physics performances of the two cores: neutron flux and spectrum, reactivity of the compensation find safety rods, neutrons balance, specific power, effective fraction of delayed neutrons, lifetime of the prompt neutrons, reactivity coefficient. We describe the hydraulic studies and experiments which have been done concerning the two cores. We discuss the criteria adopted as basis for the flow calculations. We give the results of pressure drop and sub-assembly lifting, force measurements, and vibration and pin flow distribution experiments. We discuss the constants utilized for the thermal calculations and we give the temperatures of sodium and alloy or oxide fuel, the temperature increases due to the hot points, and the limitation of the oxide fuel burn-up, originated by the pressure of the fission gases. We treat the hypotheses having been utilized for the dynamics calculations and we describe the different accidents which have been studied. We give the results of the calculations for every accident and each fuel, and we show fuel melting or sodium boiling can be avoided, even in case of the most pessimistic hypotheses, by modifying reactor characteristics (shim-rod reactivity or power of the reactor with only one cooling circuit). The reactor stability has been evaluated with the hypotheses utilized for the dynamics calculations, except of the Doppler coefficient which was intentionally increased. We show that the alloy and oxide cores are stable for every envisaged reactor power. (authors) [fr

  7. Core physics design calculation of mini-type fast reactor based on Monte Carlo method

    International Nuclear Information System (INIS)

    He Keyu; Han Weishi

    2007-01-01

    An accurate physics calculation model has been set up for the mini-type sodium-cooled fast reactor (MFR) based on MCNP-4C code, then a detailed calculation of its critical physics characteristics, neutron flux distribution, power distribution and reactivity control has been carried out. The results indicate that the basic physics characteristics of MFR can satisfy the requirement and objectives of the core design. The power density and neutron flux distribution are symmetrical and reasonable. The control system is able to make a reliable reactivity balance efficiently and meets the request for long-playing operation. (authors)

  8. Multi-physics design and analyses of long life reactors for lunar outposts

    Science.gov (United States)

    Schriener, Timothy M.

    event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete

  9. Physics design of advanced steady-state tokamak reactor A-SSTR2

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Ushigusa, Kenkichi

    2000-10-01

    Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

  10. Measurement of the physics properties of gas-cooled fast reactors in the zero energy reactor PROTEUS and analysis of the results

    International Nuclear Information System (INIS)

    Richmond, R.

    1982-12-01

    The main aim of the fast reactor physics measurements carried out in the zero energy reactor PROTEUS was to check the performance of data sets and calculation methods used in the design of fast breeder reactors. This allowed the accuracy of the power reactor calculations to be determined and enabled an assessment to be made of whether this accuracy would be sufficient to allow the design, construction and licensing of the GCFR power reactor. In order to carry out the physics measurements an existing zero energy reactor was converted to a form in which a central fast reactor lattice was surrounded by thermal zones to drive the reactor critical. One of the most important measuring techniques used to check the performance of data sets and calculation methods was the determination of reaction rate ratios and, by using an appropriate range of nuclides, it was possible to obtain a detailed picture covering 70% of reactions taking place in the central part of the fast reactor zone and with an accuracy of +-1.5% in a typical ratio. A further technique used during the work on GCFR-PROTEUS was the measurement of neutron spectrum which was carried out in a wide range of environments and, in the later stages of the work, covered the energy range from 9 keV to 2.3 MeV. These measurements, in particular, indicated significant errors in the FGL4 scattering cross-sections. A third technique, which was developed to a high degree of accuracy, was the measurement of reactivity worths. This was used in measurements of the worths of small samples and also in the application of the null reactivity technique to determine k-infinity and hence the absorption cross-sections of reactor structural materials. (Auth.)

  11. Diversity Profile of Microbes Associated with Anaerobic Sulfur Oxidation in an Upflow Anaerobic Sludge Blanket Reactor Treating Municipal Sewage

    Science.gov (United States)

    Aida, Azrina A.; Kuroda, Kyohei; Yamamoto, Masamitsu; Nakamura, Akinobu; Hatamoto, Masashi; Yamaguchi, Takashi

    2015-01-01

    We herein analyzed the diversity of microbes involved in anaerobic sulfur oxidation in an upflow anaerobic sludge blanket (UASB) reactor used for treating municipal sewage under low-temperature conditions. Anaerobic sulfur oxidation occurred in the absence of oxygen, with nitrite and nitrate as electron acceptors; however, reactor performance parameters demonstrated that anaerobic conditions were maintained. In order to gain insights into the underlying basis of anaerobic sulfur oxidation, the microbial diversity that exists in the UASB sludge was analyzed comprehensively to determine their identities and contribution to sulfur oxidation. Sludge samples were collected from the UASB reactor over a period of 2 years and used for bacterial 16S rRNA gene-based terminal restriction fragment length polymorphism (T-RFLP) and next-generation sequencing analyses. T-RFLP and sequencing results both showed that microbial community patterns changed markedly from day 537 onwards. Bacteria belonging to the genus Desulforhabdus within the phylum Proteobacteria and uncultured bacteria within the phylum Fusobacteria were the main groups observed during the period of anaerobic sulfur oxidation. Their abundance correlated with temperature, suggesting that these bacterial groups played roles in anaerobic sulfur oxidation in UASB reactors. PMID:25817585

  12. Media arrangement impacts cell growth in anaerobic fixed-bed reactors treating sugarcane vinasse: Structured vs. randomic biomass immobilization.

    Science.gov (United States)

    de Aquino, Samuel; Fuess, Lucas Tadeu; Pires, Eduardo Cleto

    2017-07-01

    This study reports on the application of an innovative structured-bed reactor (FVR) as an alternative to conventional packed-bed reactors (PBRs) to treat high-strength solid-rich wastewaters. Using the FVR prevents solids from accumulating within the fixed-bed, while maintaining the advantages of the biomass immobilization. The long-term operation (330days) of a FVR and a PBR applied to sugarcane vinasse under increasing organic loads (2.4-18.0kgCODm -3 day -1 ) was assessed, focusing on the impacts of the different media arrangements over the production and retention of biomass. Much higher organic matter degradation rates, as well as long-term operational stability and high conversion efficiencies (>80%) confirmed that the FVR performed better than the PBR. Despite the equivalent operating conditions, the biomass growth yield was different in both reactors, i.e., 0.095gVSSg -1 COD (FVR) and 0.066gVSSg -1 COD (PBR), indicating a clear control of the media arrangement over the biomass production in fixed-bed reactors. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

  14. Physical-chemical and operational performance of an anaerobic baffled reactor (ABR treating swine wastewater = Desempenho físico-químico e operacional de um reator anaeróbio compartimentado (RAC como tratamento biológico preliminar de efluentes de suinocultura

    Directory of Open Access Journals (Sweden)

    Erlon Lopes Pereira

    2010-10-01

    Full Text Available Since hog raising concentrates a huge amount of swine manure in smallareas, it is considered by the environmental government organizations to be one of the most potentially pollutant activities. Therefore the main objective of this research was to evaluate by operational criteria and removal efficiency, the performance of a Anaerobic Baffled Reactor (ABR, working as a biological pre-treatment of swine culture effluents. The physical-chemical analyses carried out were: total COD, BOD5, total solids (TS, fix (TFS and volatiles (TVS, temperature, pH, total Kjeldahl nitrogen, phosphorus, total acidity and alkalinity. The ABR unit worked with an average efficiency of 65.2 and 76.2%, respectively, concerning total COD and BOD5, with a hydraulic retention time (HRT about 15 hours. The results for volumetric organic loading rate (VOLR, organic loading rate (OLR andhydraulic loading rate (HLR were: 4.46 kg BOD m-3 day-1; 1.81 kg BOD5 kg TVS-1 day-1 and 1.57 m3 m-3 day-1, respectively. The average efficiency of the whole treatment system for total COD and BOD5 removal were 66.5 and 77.8%, showing an adequate performancein removing the organic matter from swine wastewater.A suinocultura por ser uma atividade pecuária concentradora de dejetos em pequenas áreas é considerada, pelos órgãos de gerência ambiental, como uma das atividades mais degradadoras do meio ambiente. Nesta pesquisa objetivou-se, por conseguinte, avaliar a utilização de um reator anaeróbio compartimentado (RAC, como unidade de prétratamento de um reator tipo UASB, em escala piloto, na adequação ambiental dos efluentes de suinocultura, avaliando critérios operacionais e a eficiência. As análises físico-químicasrealizadas foram: DQOtotal, DBO5, sólidos totais (ST, fixos (SF e voláteis (SV, temperatura, pH, nitrogênio total Kjeldahl, fósforo, acidez total e alcalinidade. A unidade RAC trabalhou com eficiência de 65,2 e 76,2% para a remoção de DQOtotal e DBO5

  15. Tokamak reactor for treating fertile material or waste nuclear by-products

    Science.gov (United States)

    Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

    2012-10-02

    Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

  16. Coagulant recovery from water treatment plant sludge and reuse in post-treatment of UASB reactor effluent treating municipal wastewater.

    Science.gov (United States)

    Nair, Abhilash T; Ahammed, M Mansoor

    2014-09-01

    In the present study, feasibility of recovering the coagulant from water treatment plant sludge with sulphuric acid and reusing it in post-treatment of upflow anaerobic sludge blanket (UASB) reactor effluent treating municipal wastewater were studied. The optimum conditions for coagulant recovery from water treatment plant sludge were investigated using response surface methodology (RSM). Sludge obtained from plants that use polyaluminium chloride (PACl) and alum coagulant was utilised for the study. Effect of three variables, pH, solid content and mixing time was studied using a Box-Behnken statistical experimental design. RSM model was developed based on the experimental aluminium recovery, and the response plots were developed. Results of the study showed significant effects of all the three variables and their interactions in the recovery process. The optimum aluminium recovery of 73.26 and 62.73 % from PACl sludge and alum sludge, respectively, was obtained at pH of 2.0, solid content of 0.5 % and mixing time of 30 min. The recovered coagulant solution had elevated concentrations of certain metals and chemical oxygen demand (COD) which raised concern about its reuse potential in water treatment. Hence, the coagulant recovered from PACl sludge was reused as coagulant for post-treatment of UASB reactor effluent treating municipal wastewater. The recovered coagulant gave 71 % COD, 80 % turbidity, 89 % phosphate, 77 % suspended solids and 99.5 % total coliform removal at 25 mg Al/L. Fresh PACl also gave similar performance but at higher dose of 40 mg Al/L. The results suggest that coagulant can be recovered from water treatment plant sludge and can be used to treat UASB reactor effluent treating municipal wastewater which can reduce the consumption of fresh coagulant in wastewater treatment.

  17. Progression of the lattice physics concept for the Canadian Supercritical Water Reactor

    International Nuclear Information System (INIS)

    Pencer, J.; Colton, A.

    2013-01-01

    The Canadian Supercritical Water Reactor (SCWR) is a GEN-IV reactor concept with features that support enhanced safety, clean energy, sustainability, economics and non-proliferation. Development of the lattice and core physics concepts for the SCWR has therefore focused on these features, with particular emphasis on safety and sustainability. Recently, a new two-ring fuel concept was adopted in combination with a central flow tube in the fuel channel. The combination of these two features leads to an approximately 40% increase in exit burnup and guarantees negative coolant void reactivity throughout the operating cycle. The progression from earlier concepts to the present physics concept are discussed and reviewed in this paper. (author)

  18. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  19. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    OpenAIRE

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  20. Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor

    International Nuclear Information System (INIS)

    Karpov, V.A.; Protsenko, A.N.

    1975-01-01

    Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)

  1. Proceeding on the scientific meeting and presentation on accelerator technology and its applications: physics, nuclear reactor

    International Nuclear Information System (INIS)

    Pramudita Anggraita; Sudjatmoko; Darsono; Tri Marji Atmono; Tjipto Sujitno; Wahini Nurhayati

    2012-01-01

    The scientific meeting and presentation on accelerator technology and its applications was held by PTAPB BATAN on 13 December 2011. This meeting aims to promote the technology and its applications to accelerator scientists, academics, researchers and technology users as well as accelerator-based accelerator research that have been conducted by researchers in and outside BATAN. This proceeding contains 23 papers about physics and nuclear reactor. (PPIKSN)

  2. Hydraulic retention time impact of treated recirculated leachate on the hydrolytic kinetic rate of coffee pulp in an acidogenic reactor.

    Science.gov (United States)

    Houbron, E; González-López, G I; Cano-Lozano, V; Rustrían, E

    2008-01-01

    This study attempted to investigate the impact of HRT of treated leachate recirculation on hydrolysis solubilization rate of coffee pulp in an acidogenic reactor. Coffee pulp presents more than 70% of organic matter and around of 30% of lignin and cellulose. Five lab scale reactors of 20 litres were used. Each reactor was fed with 5 kg of fresh coffee pulp and anaerobic sludge was used as inoculate. HRT of 0.5, 1, 3 and 10 days were applied. Each experiment shows that Total, Soluble and VFA COD appear rapidly in the removed leachate. HRT have a great impact on hydrolytic rate with an optimal value of 32,000 mg x L(-1) x d(-1).Low HRT increases hydrolysis rate and in consequence reduces duration of the hydrolytic phase. Also composition and concentration of VFA are influenced by HRT. Low ones favour acetic acid production and high ones permit the production of butyric. Low HRT generates leachate more easily fermentable. Efficiency of solubilization and acidification are independent of the HRT and present average values of 78% and 65% respectively. By batch feeding solid and continuous recirculation of treated leachate, HRT and SRT could be dissociated, where solid had a very high retention without problems of load, mixing and inhibition, and liquid could be recirculated with a very high rate. Under these low HRT condition, the first reactor of a two stage anaerobic system could reduces the hydrolysis duration of organic solid waste like coffee pulp and generate an optimal leachate for the methanization process. Copyright IWA Publishing 2008.

  3. A bibliography on finite element and related methods analysis in reactor physics computations (1971--1997)

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, D.C.

    1998-01-01

    This bibliography provides a list of references on finite element and related methods analysis in reactor physics computations. These references have been published in scientific journals, conference proceedings, technical reports, thesis/dissertations and as chapters in reference books from 1971 to the present. Both English and non-English references are included. All references contained in the bibliography are sorted alphabetically by the first author`s name and a subsort by date of publication. The majority of the references relate to reactor physics analysis using the finite element method. Related topics include the boundary element method, the boundary integral method, and the global element method. All aspects of reactor physics computations relating to these methods are included: diffusion theory, deterministic radiation and neutron transport theory, kinetics, fusion research, particle tracking in finite element grids, and applications. For user convenience, many of the listed references have been categorized. The list of references is not all inclusive. In general, nodal methods were purposely excluded, although a few references do demonstrate characteristics of finite element methodology using nodal methods (usually as a non-conforming element basis). This area could be expanded. The author is aware of several other references (conferences, thesis/dissertations, etc.) that were not able to be independently tracked using available resources and thus were not included in this listing.

  4. Advanced multi-physics simulation capability for very high temperature reactors

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Tak, Nam Il; Jo Chang Keun; Noh, Jae Man; Cho, Bong Hyun; Cho, Jin Woung; Hong, Ser Gi

    2012-01-01

    The purpose of this research is to develop methodologies and computer code for high-fidelity multi-physics analysis of very high temperature gas-cooled reactors(VHTRs). The research project was performed through Korea-US I-NERI program. The main research topic was development of methodologies for high-fidelity 3-D whole core transport calculation, development of DeCART code for VHTR reactor physics analysis, generation of VHTR specific 190-group cross-section library for DeCART code, development of DeCART/CORONA coupled code system for neutronics/thermo-fluid multi-physics analysis, and benchmark analysis against various benchmark problems derived from PMR200 reactor. The methodologies and the code systems will be utilized a key technologies in the Nuclear Hydrogen Development and Demonstration program. Export of code system is expected in the near future and the code systems developed in this project are expected to contribute to development and export of nuclear hydrogen production system

  5. Study on Reactor Physics Characteristic of the PWR Core Using UO2

    International Nuclear Information System (INIS)

    Tukiran Surbakti

    2009-01-01

    Study on reactor physics characteristic of the PWR core using UO 2 fuel it is necessary to be done to know the characteristic of geometry, condition and configuration of pin cell in the fuel assembly Because the geometry, configuration and condition of the pin cell in fuel core determine the loading strategy of in-core fuel management Calculation of k e ff is a part of the neutronic core parameter calculation to know the reactor physics characteristic. Generally, core calculation is done using computer code starts from modelling one unit fuel lattice cell, fuel assembly, reflector, irradiation facility and until core reactor. In this research, the modelling of pin cell and fuel assembly of the PWR 17 ×17 is done homogeneously. Calculation of the k-eff is done with variation of the fuel volume fraction, fuel pin diameter, fuel enrichment. The calculation is using by NITAWL and CENTRM, and then the results will be compared to KENOVI code. The result showed that the value of k e ff for pin cell and fuel assembly PWR 17 ×17 is not different significantly with homogenous and heterogenous models. The results for fuel volume fraction of 0.5; rod pitch 1.26 cm and fuel pin diameter of 9.6 mm is critical with burn up of 35,0 GWd/t. The modeling and calculation method accurately is needed to calculation the core physic parameter, but sometimes, it is needed along time to calculate one model. (author)

  6. BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Tsibulya, Anatoly; Rozhikhin, Yevgeniy

    2012-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. Data provided by these two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades The Russian Federation has been a major contributor to both projects with the Institute of Physics and Power Engineering (IPPE) as the major contributor from the Russian Federation. Included in the benchmark specifications from the BFS facilities are 34 critical configurations from BFS-49, 61, 62, 73, 79, 81, 97, 99, and 101; spectral characteristics measurements from BFS-31, 42, 57, 59, 61, 62, 73, 97, 99, and 101; reactivity effects measurements from BFS-62-3A; reactivity coefficients and kinetics measurements from BFS-73; and reaction rate measurements from BFS-42, 61, 62, 73, 97, 99, and 101.

  7. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto

    2017-01-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10 13 combinations and 10 11 great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  8. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2017-11-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  9. An evaluation of a mesophilic reactor for treating wastewater from a ...

    African Journals Online (AJOL)

    An evaluation of anaerobic treatment of potato-processing wastewater using an up flow Anaerobic Sludge Bed (UASB) reactor at 37°C was conducted. Wastewater from a potato-processing plant in Harare, with an average of 6.8 g COD/l, (COD = chemical oxygen demand) a high concentration of total solids (up to 6725 ...

  10. Rapid restoration of methanogenesis in an acidified UASB reactor treating 2,4,6-trichlorophenol (TCP).

    Science.gov (United States)

    Díaz-Báez, María Consuelo; Valderrama-Rincon, Juan Daniel

    2017-02-15

    Anaerobic bioreactors are often used for removal of xenobiotic and highly toxic pollutants from wastewater. Most of the time, the pollutant is so toxic that the stability of the reactor becomes compromised. It is well known that methanogens are one of the most sensitive organisms in the anaerobic consortia and hence the stability of the reactors is highly dependant on methanogenesis. Unfortunately few studies have focused on recovering the methanogenic activity once it has been inhibited by highly toxic pollutants. Here we establish a quick recovery strategy for neutralization of an acidified UASB reactor after failure by intoxication with an excess of TCP in the influent. Once the reactor returned to pH values compatible with methanogenesis, biogas production was re-started after one day and the system was re-acclimated to TCP. Successful removal of TCP from synthetic wastewater was shown for concentrations up to 70mg/L after restoration. Copyright © 2016 Elsevier B.V. All rights reserved.

  11. Microbial diversity in a full-scale anaerobic reactor treating high ...

    African Journals Online (AJOL)

    DR. NJ TONUKARI

    2012-03-22

    Mar 22, 2012 ... Microbial characteristics in the up-flow anaerobic sludge blanket reactor (UASB) of a full-scale high concentration cassava alcohol wastewater plant capable of anaerobic hydrocarbon removal were analyzed using cultivation-independent molecular methods. Forty-five bacterial operational taxonomic.

  12. Microbial diversity in a full-scale anaerobic reactor treating high ...

    African Journals Online (AJOL)

    Microbial characteristics in the up-flow anaerobic sludge blanket reactor (UASB) of a full-scale high concentration cassava alcohol wastewater plant capable of anaerobic hydrocarbon removal were analyzed using cultivation-independent molecular methods. Forty-five bacterial operational taxonomic units (OTUs) and 24 ...

  13. Effect of temperature on two-phase anaerobic reactors treating slaughterhouse wastewater

    Directory of Open Access Journals (Sweden)

    Simone Beux

    2007-11-01

    Full Text Available The effectiveness of the anaerobic treatment of effluent from a swine and bovine slaughterhouse was assessed in two sets of two-phase anaerobic digesters, operated with or without temperature control. Set A, consisting of an acidogenic reactor with recirculation and an upflow biological filter as the methanogenic phase, was operated at room temperature, while set B, consisting of an acidogenic reactor without recirculation and an upflow biological filter as the methanogenic phase, was maintained at 32°C. The methanogenic reactors showed COD (Chemical Demand of Oxygen removal above 60% for HRT (Hydraulic Retention Time values of 20, 15, 10, 8, 6, 4, and 2 days. When the HRT value in those reactors was changed to 1 day, the COD percentage removal decreased to 50%. The temperature variations did not have harmful effects on the performance of reactors in set A.Avaliou-se a eficiência do tratamento anaeróbio de efluente de matadouro de suínos e bovinos em dois conjuntos de biodigestores anaeróbios de duas fases, operados com e sem controle de temperatura. O conjunto A, formado por um reator acidogênico com recirculação e um filtro biológico de fluxo ascendente, foi operado a temperatura ambiente e o conjunto B, formado por um reator de fluxo ascendente e um filtro biológico de fluxo ascendente, foi mantido a 32°C. Os reatores metanogênicos apresentaram remoção de DQO acima de 60 % para os TRHs de 20, 15, 10, oito, seis, quatro e dois dias. Quando o TRH destes reatores foi mudado para um dia observou-se uma queda da porcentagem de remoção de DQO para 50 %. As variações de temperatura parecem não ter prejudicado o desempenho dos reatores do conjunto A.

  14. Reactor physics data for safety analysis of CANFLEX-NU CANDU-6 core

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Suk, Ho Chun

    2001-08-01

    This report contains the reactor physics data for safety analysis of CANFLEX-NU fuel CANDU-6 core. First, the physics parameters for time-average core have been described, which include the channel power and maximum bundle power map, channel axial power shape and bundle burnup. And, next the data for fuel performance such as relative ring power distribution and bundle burnup conversion ratio are represented. The transition core data from 0 to 900 full power day are represented by 100 full power day interval. Also, the data for reactivity devices of time-average core and 300 full power day of transition core are given

  15. Physical properties of heat-treated rattan waste binderless particleboard

    Science.gov (United States)

    Tajuddin, Maisarah; Ahmad, Zuraida; Halim, Zahurin; Maleque, Md Abd; Ismail, Hanafi; Sarifuddin, Norshahida

    2017-07-01

    The objective of this study is to investigate the effects of heat treatment on the properties of binderless particleboard (BPB) fabricated via hot-pressing process with pressing temperature, pressing time and pressing pressure of 180°C, 5 minutes and 1 MPa, respectively. The fabricated BPB with density in the range of 0.8-0.95g cm-3 was heated in a temperature-controlled laboratory chamber at 80°C, 120°C and 160°C for period of 2 and 8 hours before underwent physical observation, mass loss measurement and thickness swelling test. The samples had remarkable color changes, mainly with samples of treatment temperature of 160˚C, where the color differences were 9.5 and 20.3. This changed the fabricated BPB samples from yellowish brown to dark brown color when treatment conditions increased. Darker color indicates greater mass loss due to severity of chemical component in the powder. Dimensional stability of fabricated BPB was improved with higher treatment temperature as more cellulose cross-linked and hemicellulose degraded that removed the hygroscopicity behavior of powder. These results revealed that heat treatment helped in improving the BPB physical properties, particularly in dimensional stability of boards.

  16. An autotrophic nitrogen removal process: short-cut nitrification combined with ANAMMOX for treating diluted effluent from an UASB reactor fed by landfill leachate.

    Science.gov (United States)

    Liu, Jie; Zuo, Jian'e; Yang, Yang; Zhu, Shuquan; Kuang, Sulin; Wang, Kaijun

    2010-01-01

    A combined process consisting of a short-cut nitrification (SN) reactor and an anaerobic ammonium oxidation upflow anaerobic sludge bed (ANAMMOX) reactor was developed to treat the diluted effluent from an upflow anaerobic sludge bed (UASB) reactor treating high ammonium municipal landfill leachate. The SN process was performed in an aerated upflow sludge bed (AUSB) reactor (working volume 3.05 L), treating about 50% of the diluted raw wastewater. The ammonium removal efficiency and the ratio of NO2- -N to NOx- -N in the effluent were both higher than 80%, at a maximum nitrogen loading rate of 1.47 kg/(m3 x ay). The ANAMMOX process was performed in an UASB reactor (working volume 8.5 L), using the mix of SN reactor effluent and diluted raw wastewater at a ratio of 1:1. The ammonium and nitrite removal efficiency reached over 93% and 95%, respectively, after 70-day continuous operation, at a maximum total nitrogen loading rate of 0.91 kg/(m3 x day), suggesting a successful operation of the combined process. The average nitrogen loading rate of the combined system was 0.56 kg/(m3 x day), with an average total inorganic nitrogen removal efficiency 87%. The nitrogen in the effluent was mostly nitrate. The results provided important evidence for the possibility of applying SN-ANAMMOX after UASB reactor to treat municipal landfill leachate.

  17. Chemical and physical analysis of core materials for advanced high temperature reactors with process heat applications

    International Nuclear Information System (INIS)

    Nickel, H.

    1985-08-01

    Various chemical and physical methods for the analysis of structural materials have been developed in the research programmes for advanced high temperature reactors. These methods are discussed using as examples the structural materials of the reactor core - the fuel elements consisting of coated particles in a graphite matrix and the structural graphite. Emphasis is given to the methods of chemical analysis. The composition of fuel kernels is investigated using chemical analysis methods to determine the heavy metals content (uranium, plutonium, thorium and metallic impurity elements) and the amount of non-metallic constituents. The properties of the pyrocarbon and silicon carbide coatings of fuel elements are investigated using specially developed physiochemical methods. Regarding the irradiation behaviour of coated particles and fuel elements, methods have been developed for examining specimens in hot cells following exposures under reactor operating conditions, to supplement the measurements of in-reactor performance. For the structural graphite, the determination of impurities is important because certain impurities may cause pitting corrosion during irradiation. The localized analysis of very low impurity concentrations is carried out using spectrochemical d.c. arc excitation, local laser and inductively coupled plasma methods. (orig.)

  18. Exploring Stochastic Sampling in Nuclear Data Uncertainties Assessment for Reactor Physics Applications and Validation Studies

    Directory of Open Access Journals (Sweden)

    Alexander Vasiliev

    2016-12-01

    Full Text Available The quantification of uncertainties of various calculation results, caused by the uncertainties associated with the input nuclear data, is a common task in nuclear reactor physics applications. Modern computation resources and improved knowledge on nuclear data allow nowadays to significantly advance the capabilities for practical investigations. Stochastic sampling is the method which has received recently a high momentum for its use and exploration in the domain of reactor design and safety analysis. An application of a stochastic sampling based tool towards nuclear reactor dosimetry studies is considered in the given paper with certain exemplary test evaluations. The stochastic sampling not only allows the input nuclear data uncertainties propagation through the calculations, but also an associated correlation analysis performance with no additional computation costs and for any parameters of interest can be done. Thus, an example of assessment of the Pearson correlation coefficients for several models, used in practical validation studies, is shown here. As a next step, the analysis of the obtained information is proposed for discussion, with focus on the systems similarities assessment. The benefits of the employed method and tools with respect to practical reactor dosimetry studies are consequently outlined.

  19. HTGR reactor physics, thermal-hydraulics and depletion uncertainty analysis: a proposed IAEA coordinated research project

    International Nuclear Information System (INIS)

    Tyobeka, Bismark; Reitsma, Frederik; Ivanov, Kostadin

    2011-01-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis and uncertainty analysis methods. In order to benefit from recent advances in modeling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Uncertainty and sensitivity studies are an essential component of any significant effort in data and simulation improvement. In February 2009, the Technical Working Group on Gas-Cooled Reactors recommended that the proposed IAEA Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modeling be implemented. In the paper the current status and plan are presented. The CRP will also benefit from interactions with the currently ongoing OECD/NEA Light Water Reactor (LWR) UAM benchmark activity by taking into consideration the peculiarities of HTGR designs and simulation requirements. (author)

  20. Emissions of organics from bioslurry reactors treating soil contaminated with wood preserving waste

    International Nuclear Information System (INIS)

    Lewis, R.F.; Smith, M.; Hessling, J.; Dosani, M.

    1992-01-01

    This paper is a part of the work conducted for a joint Superfund Innovative Technology Evaluation (SITE) project and a study for the EPA's Office of Solid Waste and Emergency Response (OSWER) that is developing information for Best Demonstrated Available Technology (BDAT). The project was conducted at the US EPA Test and Evaluation Center located at the Gest Street Waste Water Treatment Plant in Cincinnati, Ohio. The contaminated soil chosen for the test of the effectiveness of bioslurry reactors for the degradation of wood preserving wastes was a soil from the Burlington Northern NPL site in Brainerd, Minnesota. The overall results of the soil treatment are presented in a paper titled Slurry Reactor Bioremediation of Soil-Bound Polycyclic Aromatic Hyrocarbon by Alan Jones, Madonna Brinkmann, and William Mahaffey of Ecova Corporation. Air sampling was conducted to characterize the off-gases emitted from the bioreactors during the operations and to determine organic constituent loss through volatilization. 1 tab

  1. Microbial Community Composition and Dynamics of Moving Bed Biofilm Reactor Systems Treating Municipal Sewage

    OpenAIRE

    Biswas, Kristi; Turner, Susan J.

    2012-01-01

    Moving bed biofilm reactor (MBBR) systems are increasingly used for municipal and industrial wastewater treatment, yet in contrast to activated sludge (AS) systems, little is known about their constituent microbial communities. This study investigated the community composition of two municipal MBBR wastewater treatment plants (WWTPs) in Wellington, New Zealand. Monthly samples comprising biofilm and suspended biomass were collected over a 12-month period. Bacterial and archaeal community comp...

  2. Feasibility of an Anaerobic Baffled Reactor (ABR In Treating Starch Industry Wastewater

    Directory of Open Access Journals (Sweden)

    Ali Assadi

    2007-03-01

    Full Text Available The anaerobic baffled reactor (ABR includes a mixed anaerobic culture separated into compartments and a novel process with a series of vertical baffles at each compartment. It dose not require granulation for its operation, resulting in shorter start-up time. In this study, the feasibility of the ABR process was investigated for the treatment of wheat flour starch wastewater. Simple gravity settling was used to remove suspended solids from the starch wastewater and used as feed. Start-up of a reactor (13.5L with five compartments using a diluted feed of approximately 4500 mg/L chemical oxygen demand (COD was accomplished in about 9 weeks using seed sludge from the anaerobic digester of a municipal wastewater treatment plant. The reactor with a hydraulic retention time (HRT of 72 h at 35°C and an initial organic loading rate (OLR of 1.2 kgCOD/m3.d showed a removal efficiency of 61% COD. The best reactor performance was observed with an organic loading rate of 2.5 kgCOD/m3.d (or hydraulic retention time of 2.45 d when a COD conversion of 67% was achieved. The main advantage of using an ABR comes from its compartmentalized structure. The first compartment of an ABR may act as a buffer zone to all toxic and inhibitory materials in the feed and, thus, allows the later compartments to be loaded with a relatively harmless, more uniform, and mostly acidified influent. In this respect, the later compartments would be more likely to support active populations of the relatively sensitive methanogenic bacteria.

  3. High-throughput profiling of microbial community structures in an ANAMMOX-UASB reactor treating high-strength wastewater.

    Science.gov (United States)

    Cao, Shenbin; Du, Rui; Li, Baikun; Ren, Nanqi; Peng, Yongzhen

    2016-07-01

    In this study, the microbial community structure was assessed in an anaerobic ammonium oxidation-upflow anaerobic sludge blanket (ANAMMOX-UASB) reactor treating high-strength wastewater (approximately 700 mg N L(-1) in total nitrogen) by employing Illumina high-throughput sequencing analysis. The reactor was started up and reached a steady state in 26 days by seeding mature ANAMMOX granules, and a high nitrogen removal rate (NRR) of 2.96 kg N m(-3) day(-1) was obtained at 13.2∼17.6 °C. Results revealed that the abundance of ANAMMOX bacteria increased during the operation, though it occupied a low proportion in the system. The phylum Planctomycetes was only 8.39 % on day 148 and Candidatus Brocadia was identified as the dominant ANAMMOX species with a percentage of 2.70 %. The phylum of Chloroflexi, Bacteroidetes, and Proteobacteria constituted a percentage up to 70 % in the community, of which the Chloroflexi and Bacteroidetes were likely to be related to the sludge granulation. In addition, it was found that heterotrophic denitrifying bacteria of Denitratisoma belonging to Proteobacteria phylum occupied a large proportion (22.1∼23.58 %), which was likely caused by the bacteria lysis and decay with the internal carbon source production. The SEM images also showed that plenty of other microorganisms existed in the ANAMMOX-UASB reactor.

  4. Integration of kinetic modeling and desirability function approach for multi-objective optimization of UASB reactor treating poultry manure wastewater.

    Science.gov (United States)

    Yetilmezsoy, Kaan

    2012-08-01

    An integrated multi-objective optimization approach within the framework of nonlinear regression-based kinetic modeling and desirability function was proposed to optimize an up-flow anaerobic sludge blanket (UASB) reactor treating poultry manure wastewater (PMW). Chen-Hashimoto and modified Stover-Kincannon models were applied to the UASB reactor for determination of bio-kinetic coefficients. A new empirical formulation of volumetric organic loading rate was derived for the first time for PMW to estimate the dimensionless kinetic parameter (K) in the Chen-Hashimoto model. Maximum substrate utilization rate constant and saturation constant were predicted as 11.83 g COD/L/day and 13.02 g COD/L/day, respectively, for the modified Stover-Kincannon model. Based on four process-related variables, three objective functions including a detailed bio-economic model were derived and optimized by using a LOQO/AMPL algorithm, with a maximum overall desirability of 0.896. The proposed optimization scheme demonstrated a useful tool for the UASB reactor to optimize several responses simultaneously. Copyright © 2012 Elsevier Ltd. All rights reserved.

  5. Kinetic parameters of biomass growth in a UASB reactor treating wastewater from coffee wet processing (WCWP

    Directory of Open Access Journals (Sweden)

    Claudio Milton Montenegro Campos

    2014-10-01

    Full Text Available This study evaluated the treatment of wastewater from coffee wet processing (WCWP in an anaerobic treatment system at a laboratory scale. The system included an acidification/equalization tank (AET, a heat exchanger, an Upflow Anaerobic Sludge Blanket Reactor (UASB, a gas equalization device and a gas meter. The minimum and maximum flow rates and volumetric organic loadings rate (VOLR were 0.004 to 0.037 m 3 d -1 and 0.14 to 20.29 kgCOD m -3 d -1 , respectively. The kinetic parameters measured during the anaerobic biodegradation of the WCWP, with a minimal concentration of phenolic compounds of 50 mg L - ¹, were: Y = 0.37 mgTVS (mgCODremoved -1 , Kd = 0.0075 d-1 , Ks = 1.504mg L -1 , μmax = 0.2 d -1 . The profile of sludge in the reactor showed total solids (TS values from 22,296 to 55,895 mg L -1 and TVS 11,853 to 41,509 mg L -1 , demonstrating a gradual increase of biomass in the reactor during the treatment, even in the presence of phenolic compounds in the concentration already mentioned.

  6. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  7. Reactor

    International Nuclear Information System (INIS)

    Evans, R.M.

    1976-01-01

    Disclosed is 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. 1 claim, 16 figures

  8. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

  9. Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

    2014-10-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

  10. Correction method for critical extrapolation of control-rods-rising during physical start-up of reactor

    International Nuclear Information System (INIS)

    Zhang Fan; Chen Wenzhen; Yu Lei

    2008-01-01

    During physical start-up of nuclear reactor, the curve got by lifting the con- trol rods to extrapolate to the critical state is often in protruding shape, by which the supercritical phenomena is led. In the paper, the reason why the curve was in protruding was analyzed. A correction method was introduced, and the calculations were carried out by the practical data used in a nuclear power plant. The results show that the correction method reverses the protruding shape of the extrapolating curve, and the risk of reactor supercritical phenomena can be reduced using the extrapolated curve got by the correction method during physical start-up of the reactor. (authors)

  11. Newly Available Reactor Physics Benchmark data in the March 2011 Edition of the IRPhEP Handbook

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; J. Blair Briggs; Jim Gulliford

    2011-06-01

    The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the data are compromised, it is unlikely that any of these measurements would be repeated in the future. The purpose of the IRPhEP is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Several new evaluations have been prepared for inclusion in the March 2011 edition of the IRPhEP Handbook.

  12. Identification of Archaeal population in the granular sludge of an UASB reactor treating sewage at low temperatures.

    Science.gov (United States)

    Gomec, Cigdem Y; Letsiou, Ioanna; Ozturk, Izzet; Eroglu, Veysel; Wilderer, Peter A

    2008-11-01

    Effect of low temperature on up-flow anaerobic sludge bed (UASB) reactor performance treating raw sewage was investigated in terms of the variations in methanogenic diversity using the 16S rRNA based Fluorescence In-Situ Hybridization (FISH) technique. The diversity of microorganisms present in the anaerobic granular sludge and the structure of the granules operated at 13 degrees C have been investigated using FISH combined with CSLM (Confocal Scanning Laser Microscopy). According to FISH results, archaeal cells representing methanogens were found intensively dominant in the bottom sampling port of the UASB reactor and acetoclastic Methanosaeta was the abundant methanogen. Other methanogens such as Methanosarcina and Methanobacterium like species were also observed. The abundance of originally mesophilic Methanosaeta-related Archaea under low temperature at all sampling days revealed the microbial adaptation to psychrophilic conditions. This might be attributed to the enzymatic alterations in Methanosaeta cells originating from seed sludge, which were exposed to sub-mesophilic temperatures at start-up and then to psychrophilic conditions during gradual decreases of temperature. According to CSLM observation, even though the sludge retained in the reactor kept its granular form as a whole, the majority of the granules had a tendency to be partly broken and they lost their rigidity when raw sewage was fed following synthetic sewage. Besides, Methanosaeta related species prevailing in seed sludge have noticeably lost their long filamentous forms and deteriorated during raw sewage feeding. Members of the order Methanobacteriales constituted the major hydrogenothrophic methanogens present in the psychrophilic UASB reactor, whereas the other hydrogenothrophic methanogens--members of the order Methanococcales and Methanogenium relatives--were absent.

  13. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    Directory of Open Access Journals (Sweden)

    Lindley Benjamin A.

    2016-01-01

    Full Text Available The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO2/PuO2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs-clad systems, particularly for current and near-term build LWRs. R&D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN and uranium silicide (U3Si2. Candidate cladding materials include advanced stainless steel (FeCrAl and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R&D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I2S-LWR, a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP Integrated Research Project (IRP is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I2S-LWR design are U3Si2 and advanced stainless steel respectively. In addition, the I2S-LWR design

  14. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    International Nuclear Information System (INIS)

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences

  15. Sources of radioactive waste from light-water reactors and their physical and chemical properties

    International Nuclear Information System (INIS)

    Bell, M.J.; Collins, J.T.

    1979-01-01

    The general physical and chemical properties of waste streams in light-water reactors (LWRs) are described. The principal mechanisms for release and the release pathways to the environment are discussed. The calculation of liquid and gaseous source terms using one of the available models is presented. These calculated releases are compared with observed releases from operating LWRs. The computerized mathematical model used is the GALE Code which is the Nuclear Regulatory Commission (NRC) staff's model for calculating source terms for effluents from LWRs (USNRC76a, USNRC76b). Programs currently being conducted at operating reactors by the NRC, Electric Power Research Institute, and various utilities to better define the characteristics of waste streams and the performance of radwaste process equipment are described

  16. Fast pyrolysis of creosote treated wood ties in a fluidized bed reactor and analytical characterization of product fractions

    International Nuclear Information System (INIS)

    Jung, Su-Hwa; Koo, Won-Mo; Kim, Joo-Sik

    2013-01-01

    A fraction of creosote treated wood ties was pyrolyzed in a pyrolysis plant equipped with a fluidized bed reactor and char-separation system at different temperatures. Analyses of each pyrolysis product, especially the oil, were carried out using a variety of analytical tools. The maximum oil yield was obtained at 458 °C with a value of 69.3 wt%. Oils obtained were easily separated into two phases, a creosote-derived fraction (CDF) and a wood-derived fraction (WDF). Major compounds of the WDF were acetic acid, furfural and levoglucosan, while the CDF was mainly composed of polycyclic aromatic hydrocarbons (PAHs), such as 1-methylnaphthalene, biphenyl, acenaphthene, dibenzofuran, fluorene, phenanthrene, anthracene, fluoranthene and pyrene. HPLC analysis showed that the concentration of PAHs of the CDF obtained at 458 °C constituted about 22.5 wt% of the oil. - Highlights: • Creosote treated wood ties was stably pyrolyzed in a fluidized bed reactor. • Pyrolysis oil contained extremely low metal content due to the char removal system. • Bio-oil components was quantitatively analyzed by relative response factor. • Creosote-derived pyrolysis oil fraction was composed of PHAs and has a high caloric value (39 MJ/kg)

  17. Study on treating of low-level radioactive reactor wastewater by combined membrane process (UF-RO)

    International Nuclear Information System (INIS)

    Lu Yunyun; Cao Qiru; Chen Yunming; Huang Lijuan; Bai Xiaofeng; Li Bing; Feng Liang

    2013-01-01

    According to the characteristics of radionuclide exists in the low-level radioactive reactor waste water from HFETR, we use a new combined membrane process separation technology to study the efficient treating of low-lever radioactive reactor wastewater. First, the prepared the simulated wastewater contained Cs + , Sr 2+ , CO 2+ , Ni 2+ , and Fe 3+ . Then, we sequentially investigated the pressure, ion concentration, pH value and EDTA, which have effects on the desalination rate of membrane processing metal ions in wastewater. The results show that: in the condition of pH = 7, and added 0.15 mol/L EDTA, the simulated wastewater separated by UF-RO, desalination rates of Cs + , Sr 2+ , CO 2+ , Ni 2+ and Fe 3+ are all above 95%; In the subsequent trials, adding 0.15 mol/L EDTA into the radioactive residuary solution, and then treating by UF-RO-RO, the decontamination efficiency can reach 95.7%. (authors)

  18. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  19. An analysis of the physical properties of recovered CCA-treated wood from residential decks

    Science.gov (United States)

    David Bailey; Robert L. Smith; Philip A. Araman

    2004-01-01

    A large volume of CCA-treated wood removed from residential decks is disposed of in landfills every year, and better environmentally conscious alternatives are needed. Recycling CCA-treated wood from the decks could be a feasible alternative, but there is a lack of knowledge regarding the physical properties of the material. This research analyzed the chemical and...

  20. Improvement of inherent safety features in CSR (Coupled Spectrum Reactor) for treating MA

    International Nuclear Information System (INIS)

    Aziz, F.; Kitamoto, Asashi.

    1996-01-01

    Burning and/or transmutation (B/T) of MA is proposed here using a CSR (Coupled Spectrum Reactor) concept. CSR was based on a modified conventional 1150 MWe-PWR system, and consisted of two core regions for thermal and fast neutrons, respectively. The B/T fuel used was supposed such that MA discharged from 1 GWe-LWR were mixed homogeneously in LWR fuel. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio, (V m /V f ). In order to improve its inherent safety features, several cases of CSR were studied and compared, each case used different fuel type in the inner region. The result of the calculations showed that safety features can be improved by using composite fuel of ( 235 U-Pu- 238 U) in the inner region. The equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute MA up to 808 kg/stage in a single reactor operated with a reactivity swing of 2.8 % Δk/kk'. (author)

  1. The microbial community of a passive biochemical reactor treating arsenic, zinc and sulfate-rich seepage

    Directory of Open Access Journals (Sweden)

    Susan Anne Baldwin

    2015-03-01

    Full Text Available Sulfidogenic biochemical reactors for metal removal that use complex organic carbon have been shown to be effective in laboratory studies, but their performance in the field is highly variable. Successful operation depends on the types of microorganisms supported by the organic matrix, and factors affecting the community composition are unknown. A molecular survey of a field-based biochemical reactor that had been removing zinc and arsenic for over six years revealed that the microbial community was dominated by methanogens related to Methanocorpusculum sp. and Methanosarcina sp., which co-occurred with Bacteroidetes environmental groups, such as Vadin HA17, in places where the organic matter was more degraded. The metabolic potential for organic matter decomposition by Ruminococcaceae was prevalent in samples with more pyrolysable carbon. Rhodobium- and Hyphomicrobium-related genera within the Rhizobiales Order that have the metabolic potential for dark hydrogen fermentation and methylotrophy, and unclassified Comamonadaceae were the dominant Proteobacteria. The unclassified environmental group Sh765B-TzT-29 was an important Delta-Proteobacteria group in this BCR, that co-occurred with the dominant Rhizobiales OTUs. Organic matter degradation is one driver for shifting the microbial community composition and therefore possibly the performance of these bioreactors over time.

  2. Long-term evaluation of a sequential batch reactor (SBR) treating dairy wastewater for carbon removal.

    Science.gov (United States)

    Gutiérrez, Soledad; Ferrari, Adrián; Benítez, Alejandra; Travers, Dayana; Menes, Javier; Etchebehere, Claudia; Canetti, Rafael

    2007-01-01

    Many dairy industries have been using SBR wastewater treatment plants because they allow optimal working condition to be reached. However, to take advantage of SBR capabilities, strong process automation is needed. The aim of this work is to study the factors that influence SBR performance to improve modelling and control. To better understand the whole process we studied the kinetic modelling, the carbon removal mechanism and the relation between reactor performance, aerobic heterotrophic activity and bacterial population dynamics (by terminal restriction fragment length polymorphisms of 16S rDNA, T-RFLP). The heterotrophic activity values presented high variability during some periods; however, this was not reflected on the reactor performance. As sludge health indicator, the average activity in a period was better than individual values. Although all the carbon removal mechanisms are still unclear for this process, they seemed to be influenced by non-respirometric ways (storage, biosorption, accumulation, etc.). The variability of heterotrophic activity could be correlated with the bacterial population diversity over time. Despite the high variability of the activity, a simple kinetic model (pseudo ASM1) based on apparent constant parameters was developed and calibrated. Such modellisation provided a good tool for control purposes.

  3. Experimental study of hydrodynamic and operation start of a baffled anaerobic reactor treating sewage

    Directory of Open Access Journals (Sweden)

    Ana Carolina Silveira Perico

    2009-12-01

    Full Text Available It is important to provide individual sanitation systems for sewage peri-urban communities or rural areas to minimize impacts on the environment and human health caused by sewage discharge in natura into water resources. In this context, the anaerobic digestion of effluent has been one of the main considered technologies due to easy implementation, material minimization and reduction in waste production. The objective of this work was to study a Baffled Anaerobic Reactor (BAR including its hydrodynamic characteristics, percentile of inoculum to be applied and reactor operation start. It was concluded that the flow is dispersed with 3.84% of dead spaces and that 20% of the cow manure provided best results; however, due to the high fiber content of the manure, its use is not recommended as inoculum. The BAR system, composed of four chambers, presented good performance for sewage treatment of a rural community in terms of organic substance removal (COD, turbidity and solids meeting effluent disposal standards of these parameters considering the Federal and Minas Gerais State legislation, in Brazil, even in a transient phase of operation, at temperatures below 20°C. However, the effluents from the BAR can’t be released into water bodies without other parameters such as nitrogen, phosphorus, fecal coliforms, and others are investigated to be conforming to those standards.

  4. Simulation of the Performance of a Fundamental Neutron Physics Beamline at the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mahurin, R. [University of Tennessee, Knoxville (UTK); Greene, Geoffrey L [ORNL; Koehler, Paul Edward [ORNL; Cianciolo, Vince [ORNL

    2005-05-01

    We study the expected performance of the proposed fundamental neutron physics beamline at the upgraded High Flux Isotope Reactor at Oak Ridge National Laboratory. A curved neutron guide transmits the neutrons from the new cold source into a guide hall. A novel feature of the proposed guide is the use of vertical focusing to increase the flux for experiments that require relatively small cross-section beams. We use the simulation code IB to model straight, multi-channel curved, and tapered guides of various m values. Guide performance for the current NPDGamma and proposed abBA experiments is evaluated.

  5. Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

    International Nuclear Information System (INIS)

    Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

    1993-12-01

    Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

  6. IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bostelmann, F. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained). SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on

  7. Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

    Science.gov (United States)

    Tereshonok, V. A.; Kryakvin, L. V.; Pitilimov, V. A.; Karpov, S. A.; Kulikov, V. I.; Zhylmaganbetov, N. M.; Kavun, O. Yu.; Popykin, A. I.; Shevchenko, R. A.; Shevchenko, S. A.; Semenova, T. V.

    2017-12-01

    The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

  8. Reactor physics teaching and research in the Swiss nuclear engineering master

    International Nuclear Information System (INIS)

    Chawla, R.

    2012-01-01

    Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

  9. On the accuracy of reactor physics calculations for square HPLWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)]. E-mail: fabian.jatuff@psi.ch; Macku, K. [Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Chawla, R. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institute, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2006-01-15

    Although the supercritical-pressure or high-performance light water reactor (HPLWR) concept is largely based on the well-established technological experience available with conventional light water reactors, there is still no consensus on various key design features such as an optimal layout for the fuel assembly. This results mainly from the very large density variations of supercritical-pressure water in the core, which render it difficult to ensure reliable values for parameters such as power peaking factors and reactivity worths. The present paper describes studies carried out to compare deterministic and Monte Carlo codes for analysing a representative square HPLWR lattice with uniform 5%-enriched UO{sub 2} fuel. The main purpose has been to assess the prediction accuracies achievable for integral parameters such as the multiplication factor, control absorber effectiveness, moderator/coolant density reactivity feedback and pin power distributions. The results show good agreement between the deterministic and stochastic calculations for the unperturbed lattice. However, for certain perturbed situations involving, for example, local coolant density changes in the assembly or control absorber insertion, the observed discrepancies are large enough to question the basic viability of the reactor physics design, e.g. with respect to the thermal performance.

  10. Physical and chemical feasibility of fueling molten salt reactors with TRU's trifluorides

    International Nuclear Information System (INIS)

    Ignatiev, V.; Feinberg, O.; Konakov, S.; Subbotine, S.; Surenkov, A.; Zakirov, R.

    2001-01-01

    The molten salt reactor (MSR) concept is very important for consideration as an element of future nuclear energy systems. These reactor systems are unique in many ways. Particularly, the MSRs appear to have substantial promise not only as advanced TRU free system operating in U-Th cycle, but also as transmuter of TRU. Physical and chemical feasibility of fueling MSR with TRU trifluorides is examined. Solvent compositions with and without U-Th as fissile / fertile addition are considered. The principle reactor and fuel cycle variables available for optimizing the performance of MSR as TRU transmuting system are discussed. These efforts led to the definition in minimal TRU mass flow rate, reduced total losses to waste and maximum possible burn up rate for the molten salt transmuter. The current status of technology and prospects for revisited interest are summarized. Significant chemical problems are remain to be resolved at the end of prior MSRs programs, notably, graphite life durability, tritium control, fate of noble metal fission products. Questions arising from plutonium and minor actinide fueling include: corrosion and container chemistry, new redox buffer for systems without uranium, analytical chemistry instrumentation, adequate constituent solubilities, suitable fuel processing and waste form development. However these problems appear to be soluble. (author)

  11. Progress report on research and development in 1991, Institute of Neutron Physics and Reactor Engineering, KfK

    International Nuclear Information System (INIS)

    1992-03-01

    Progress report on research and development in 1991 Institute of Neutron Physics and Reactor Engineering. The Institute of Neutron Physics and Reactor Engineering is concerned with research work in the field of nuclear engineering related to the safety of fast and thermal reactors as well as with specific problems of fusion reactor technology. Under the project of nuclear safety research, the Institute works on concepts designed to drastically improve reactor safety. Apart from that, methods to estimate and minimize the radiological consequences of reactor accidents are developed. Under the fusion technology project, the Institute deals with neutron physics and technological questions of the breeding blanket. Basic research covers technico-physical questions of the interaction between light ion radiation of a high energy density and matter. In addition and to a small extent, questions of employing hydrogen in the transport area are studied. For all these tasks it is indispensable to use up-to-date data processing methods and equipment, from the highest capacity computer to the integrated minicomputer system. (orig./DG) [de

  12. The New Cold Neutron Radiography Facility (CNRF) at the Mianyang Research Reactor of the China Academy of Engineering Physics

    Science.gov (United States)

    Bin, Tang; Heyong, Huo; Ke, Tang; Rogers, John; Haste, Martin; Christodoulou, Marios

    A new cold neutron radiography beamline has been designed and constructed for the Mianyang reactor at the Institute of Nuclear Physics and Chemistry of the China Academy of Engineering Physics. This paper describes the components of the system and demonstrates the achievable image resolution.

  13. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    International Nuclear Information System (INIS)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M.; Styrine, Y.A.

    2000-01-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included

  14. A Roadmap and Discussion of Issues for Physics Analyses Required to Support Plutonium Disposition in VVER-1000 Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Primm, R.T.; Drischler, J.D.; Pavlovichev, A.M. Styrine, Y.A.

    2000-06-01

    The purpose of this report is to document the physics analyses that must be performed to successfully disposition weapons-usable plutonium in VVER-1000 reactors in the Russian Federation. The report is a document to support programmatic and financial planning. It does not include documentation of the technical procedures by which physics analyses are performed, nor are the results of any analyses included.

  15. Microbial monitoring of ammonia removal in a UASB reactor treating pre-digested chicken manure with anaerobic granular inoculum.

    Science.gov (United States)

    Yangin-Gomec, Cigdem; Pekyavas, Goksen; Sapmaz, Tugba; Aydin, Sevcan; Ince, Bahar; Akyol, Çağrı; Ince, Orhan

    2017-10-01

    Performance and microbial community dynamics in an upflow anaerobic sludge bed (UASB) reactor coupled with anaerobic ammonium oxidizing (Anammox) treating diluted chicken manure digestate (Total ammonia nitrogen; TAN=123±10mg/L) were investigated for a 120-d operating period in the presence of anaerobic granular inoculum. Maximum TAN removal efficiency reached to above 80% with as low as 20mg/L TAN concentrations in the effluent. Moreover, total COD (tCOD) with 807±215mg/L in the influent was removed by 60-80%. High-throughput sequencing revealed that Proteobacteria, Actinobacteria, and Firmicutes were dominant phyla followed by Euryarchaeota and Bacteroidetes. The relative abundance of Planctomycetes significantly increased from 4% to 8-9% during the late days of the operation with decreased tCOD concentration, which indicated a more optimum condition to favor ammonia removal through anammox route. There was also significant association between the hzsA gene and ammonia removal in the UASB reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. COD fractions changes in the SBR-type reactor treating municipal wastewater with controlled percentage of dairy sewage

    Science.gov (United States)

    Struk-Sokołowska, Joanna; Rodziewicz, Joanna

    2017-11-01

    The aim of study was to investigate the influence of percentage of dairy wastewater in the municipal wastewater on the changes of COD fractions during the cycle of SBR-type reactor. The scope of the research included physicochemical analyses of municipal wastewater without dairy wastewater, dairy wastewater, mixture of municipal and dairy wastewater as well as treated sewage. Both the concentrations and the proportions between COD fractions changed in the SBR cycle. In raw municipal and dairy wastewater - XS, insoluble hardly bio-degradable fraction of COD dominated (49.6 and 64.5% respectively). In treated wastewater SI, COD for dissolved compounds that are not biologically decomposed (inert) (from 62.1 to 74.6%) dominated, while XS fraction was from 19.1 to 24.4%. The consumption rate of organic compounds depended on the type of COD fraction, SBR cycle phase and the percentage of dairy wastewater. The highest rates of organic compounds consumption were noted in the phase of mixing. In the case of fraction SI, no differences in concentration in the SBR cycle time, were found. Concentration of COD in treated wastewater was from 34.8 to 58.9 mgO2·L-1 (efficiency wastewater treatment from 96.0 to 98.6%).

  17. Performance of staged and non-staged up-flow anaerobic sludge bed (USSB and UASB) reactors treating low strength complex wastewater.

    Science.gov (United States)

    Sevilla-Espinosa, Susana; Solórzano-Campo, Maricela; Bello-Mendoza, Ricardo

    2010-09-01

    The use of anaerobic processes to treat low-strength wastewater has been increasing in recent years due to their favourable performance-costs balance. For optimal results, it is necessary to identify reactor configurations that are best suited for this kind of application. This paper reports on the comparative study carried out with two high-rate anaerobic reactor systems with the objective of evaluating their performances when used for the treatment of low-strength, complex wastewater. One of the systems is the commonly used up-flow anaerobic sludge blanket (UASB) reactor. The other is the up-flow staged sludge bed (USSB) system in which the reactor was divided longitudinally into 3, 5 and 7 compartments by the use of baffles. The reactors (9 l) were fed with a synthetic, soluble and colloidal waste (chemical oxygen demand (COD) flow hydraulics, between plug-flow and completely-mixed, in the UASB and 7 stages USSB reactors allowed efficient degradation of substrates with minimum effluent concentrations. Low number of compartments in the USSB reactors increased the levels of short-circuiting thus reducing substrate removal efficiencies. All reactors showed high COD removal efficiencies (93-98%) and thus can be regarded as suitable for the treatment of low strength, complex wastewater. Staged anaerobic reactors can be a good alternative for this kind of application provided they are fitted with a large enough (> or =7) number of compartments to fully take advantage of their strengths. Scale factors seem to have influenced importantly on the comparison between one and multi staged sludge-bed reactors and, therefore, observations made here could change at larger reactor volumes.

  18. A bustling academic reactor creates challenges and opportunities in the area of physical protection

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.; Standley, V.

    2001-01-01

    The 250 kW TRIGA research reactor is located at the Atominstitut Vienna, Austria, only a few subway stations from the city centre of Vienna. Its main purpose is the training of university students in the field of nuclear engineering and radiation protection as well as in radiochemistry and neutron- and solid-state physics. The existing facility is visited during a normal academic year by about 300 persons per day falling into seven different categories including fully employed staff, students occasionally visiting a seminar, and IAEA personnel from all over the world. These different groups have to be accounted for daily and are separated into different categories in view of security and physical protection. (author)

  19. Fast reactors fuel cycle core physics results from the CAPRA-CADRA programme

    International Nuclear Information System (INIS)

    Vasile, A.; Rimpault, G.; Tommasi, J.; Saint Jean, C. de; Delpech, M.; Hesketh, K.; Beaumont, H.M.; Sunderland, R.E.; Newton, T.; Smith, P.; Raedt, Ch. de; Vambenepe, G.; Lefevre, J.C.; Maschek, W.; Haas, D

    2001-01-01

    This paper presents an overview of fast reactor core physics results obtained in the context of the CAPRA-CADRA European collaborative programme, whose aim is to investigate a broad range of possible options for plutonium and radioactive waste management. Different types of fast reactors have been studied to evaluate their potential capabilities with respect to the long term management of plutonium, minor actinides (MAs) and long- lived fission products (LLFPs). Among the several options aiming at reducing waste and consequently radio toxicity are: homogeneous recycling of Minor Actinides, heterogeneous recycling of Minor Actinides either without or with moderation, dedicated critical cores (fuelled mainly with Minor Actinides) and Accelerator Driven System (ADS) variants. In order to achieve a detailed understanding of the potential of the various options, advanced core physics methods have been implemented and tested and applied, for example, to improving control rod modeling and to studying safety aspects. There has also been code development and experimental work carried out to improve the understanding of fuel performance behaviors. (author)

  20. The development of the nuclear physics in Latvia III. The research nuclear reactor IRT begins to work in Latvia

    International Nuclear Information System (INIS)

    Ulmanis, U.

    2005-01-01

    This article is associated with the study of reactors technical parameters with specific interest on the effect the distribution of neutron and gamma radiation through the reactor's cooling systems has on the environment. Scientist began by implementing monitoring system to assist in the research of nuclear spectroscopy, neutron activation analysis, neutron diffraction, solid-state radiation physics, chemistry and radiobiology. The first sets of results are summarized with in the article. (author)

  1. Microbial characterization and degradation of linear alkylbenzene sulfonate in an anaerobic reactor treating wastewater containing soap powder.

    Science.gov (United States)

    Carosia, Mariana Fronja; Okada, Dagoberto Yukio; Sakamoto, Isabel Kimiko; Silva, Edson Luiz; Varesche, Maria Bernadete Amâncio

    2014-09-01

    The aim of this study was to evaluate the removal of linear alkylbenzene sulfonate (LAS) in an anaerobic fluidized bed reactor (AFBR) treating wastewater containing soap powder as LAS source. At Stage I, the AFBR was fed with a synthetic substrate containing yeast extract and ethanol as carbon sources, and without LAS; at Stage II, soap powder was added to this synthetic substrate obtaining an LAS concentration of 14 ± 3 mg L(-1). The compounds of soap powder probably inhibited some groups of microorganisms, increasing the concentration of volatile fatty acids (VFA) from 91 to 143 mg HAc L(-1). Consequently, the LAS removal rate was 48 ± 10% after the 156 days of operation. By sequencing, 16S rRNA clones belonging to the phyla Proteobacteria and Synergistetes were identified in the samples taken at the end of the experiment, with a remarkable presence of Dechloromonas sp. and Geobacter sp. Copyright © 2014 Elsevier Ltd. All rights reserved.

  2. Microbial community composition and dynamics of moving bed biofilm reactor systems treating municipal sewage.

    Science.gov (United States)

    Biswas, Kristi; Turner, Susan J

    2012-02-01

    Moving bed biofilm reactor (MBBR) systems are increasingly used for municipal and industrial wastewater treatment, yet in contrast to activated sludge (AS) systems, little is known about their constituent microbial communities. This study investigated the community composition of two municipal MBBR wastewater treatment plants (WWTPs) in Wellington, New Zealand. Monthly samples comprising biofilm and suspended biomass were collected over a 12-month period. Bacterial and archaeal community composition was determined using a full-cycle community approach, including analysis of 16S rRNA gene libraries, fluorescence in situ hybridization (FISH) and automated ribosomal intergenic spacer analysis (ARISA). Differences in microbial community structure and abundance were observed between the two WWTPs and between biofilm and suspended biomass. Biofilms from both plants were dominated by Clostridia and sulfate-reducing members of the Deltaproteobacteria (SRBs). FISH analyses indicated morphological differences in the Deltaproteobacteria detected at the two plants and also revealed distinctive clustering between SRBs and members of the Methanosarcinales, which were the only Archaea detected and were present in low abundance (suspended communities from both plants were diverse and dominated by aerobic members of the Gammaproteobacteria and Betaproteobacteria. This study represents the first detailed analysis of microbial communities in full-scale MBBR systems and indicates that this process selects for distinctive biofilm and planktonic communities, both of which differ from those found in conventional AS systems.

  3. Micronutrient component changes in the biogas slurry treated by a pilot solar-heated anaerobic reactor

    Science.gov (United States)

    Yang, Z. Y.; Xu, Y. B.; Li, P. F.; Wang, Y. J.; Sun, J.; Zhang, Y. P.

    2017-06-01

    A solar-heated anaerobic reactor system was applied to decompose livestock wastewater, in which cattle manure and chopped straw were mixed (CODCr 15,000∼25,000 mg·l-1), the commercial microorganisms were added to ambient acidification (about 32°C) and the acclimated sludge was inoculated. Then, the experiments were carried out on wastewater anaerobic degradation and biogas production at 40∼42°C, as fed every 10 days till stable running. The results showed that NH3-N and PO4 3- of the biogas slurry were 441 mg·l-1 and 65.0 mg·l-1 on the 35th day, respectively. The concentration of K was up to 350 mg·l-1 in the biogas slurry, rather higher than that of Mg and Fe, which indicated that the available K could contribute more in the agricultural irrigation. Total amino acids were up to 23.7 mg·l-1 after anaerobic digestion, in which Lys, Thr, Ala and Arg were prominent in the biogas slurry. These amino acids could be beneficial to seed soaking, feed adding and apply as foliar fertilizer. The major volatile organic compounds were detected in the biogas slurry, including toluene, m-cresol (up to 0.036% in the process of ambient acidification) and triethylsilane, which could be reduced to scarcely influence on agricultural application after anaerobic digestion.

  4. Start up study of UASB reactor treating press mud for biohydrogen production

    International Nuclear Information System (INIS)

    Radjaram, B.; Saravanane, R.

    2011-01-01

    Anaerobic digestion of press mud mixed with water for biohydrogen production was performed in continuous fed UASB bioreactor for 120 days. Experiment was conducted by maintaining constant HRT of 30 h and the volume of biohydrogen evolved daily was monitored. Various parameters like COD, VFA, Alkalinity, EC, Volatile solids, pH with respect to biohydrogen production were monitored at regular interval of time. SBPR was 10.98 ml g -1 COD reduced d -1 and 12.77 ml g -1 VS reduced d -1 on peak yield of biohydrogen. COD reduction was above 70 ± 7%. Maximum gas yield was on the 78th day to 2240 ml d -1 . The aim of the experiment is to study the startup process of UASB reactor for biohydrogen production by anaerobic fermentation of press mud. The inoculum for the process is cow dung and water digested in anaerobic condition for 30 days with municipal sewage sludge. The study explores the viability of biohydrogen production from press mud which is a renewable form of energy to supplement the global energy crisis. -- Highlights: → Feasibility of biohydrogen production from press mud was explored in this study. The gas yield was maximum on the 78th day to 2240 ml d -1 with H 2 % of 52-59%. Biohydrogen yield was about 890 ml kg -1 press mud added d -1 . Press mud is identified as an excellent potential waste to tap energy.

  5. Aerobic granulation strategy for bioaugmentation of a sequencing batch reactor (SBR) treating high strength pyridine wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiaodong; Chen, Yan [Jiangsu Key Laboratory for Chemical Pollution Control and Resources Reuse, School of Environmental and Biological Engineering, Nanjing University of Science and Technology, Nanjing 210094, Jiangsu Province (China); Zhang, Xin [Jiangsu Key Laboratory for Chemical Pollution Control and Resources Reuse, School of Environmental and Biological Engineering, Nanjing University of Science and Technology, Nanjing 210094, Jiangsu Province (China); Suzhou Institute of Architectural Design Co., Ltd, Suzhou 215021, Jiangsu Province (China); Jiang, Xinbai; Wu, Shijing [Jiangsu Key Laboratory for Chemical Pollution Control and Resources Reuse, School of Environmental and Biological Engineering, Nanjing University of Science and Technology, Nanjing 210094, Jiangsu Province (China); Shen, Jinyou, E-mail: shenjinyou@mail.njust.edu.cn [Jiangsu Key Laboratory for Chemical Pollution Control and Resources Reuse, School of Environmental and Biological Engineering, Nanjing University of Science and Technology, Nanjing 210094, Jiangsu Province (China); Sun, Xiuyun; Li, Jiansheng; Lu, Lude [Jiangsu Key Laboratory for Chemical Pollution Control and Resources Reuse, School of Environmental and Biological Engineering, Nanjing University of Science and Technology, Nanjing 210094, Jiangsu Province (China); Wang, Lianjun, E-mail: wanglj@mail.njust.edu.cn [Jiangsu Key Laboratory for Chemical Pollution Control and Resources Reuse, School of Environmental and Biological Engineering, Nanjing University of Science and Technology, Nanjing 210094, Jiangsu Province (China)

    2015-09-15

    Abstract: Aerobic granules were successfully cultivated in a sequencing batch reactor (SBR), using a single bacterial strain Rhizobium sp. NJUST18 as the inoculum. NJUST18 presented as both a good pyridine degrader and an efficient autoaggregator. Stable granules with diameter of 0.5–1 mm, sludge volume index of 25.6 ± 3.6 mL g{sup −1} and settling velocity of 37.2 ± 2.7 m h{sup −1}, were formed in SBR following 120-day cultivation. These granules exhibited excellent pyridine degradation performance, with maximum volumetric degradation rate (V{sub max}) varied between 1164.5 mg L{sup −1} h{sup −1} and 1867.4 mg L{sup −1} h{sup −1}. High-throughput sequencing analysis exhibited a large shift in microbial community structure, since the SBR was operated under open condition. Paracoccus and Comamonas were found to be the most predominant species in the aerobic granule system after the system had stabilized. The initially inoculated Rhizobium sp. lost its dominance during aerobic granulation. However, the inoculation of Rhizobium sp. played a key role in the start-up process of this bioaugmentation system. This study demonstrated that, in addition to the hydraulic selection pressure during settling and effluent discharge, the selection of aggregating bacterial inocula is equally important for the formation of the aerobic granule.

  6. Aerobic granulation strategy for bioaugmentation of a sequencing batch reactor (SBR) treating high strength pyridine wastewater

    International Nuclear Information System (INIS)

    Liu, Xiaodong; Chen, Yan; Zhang, Xin; Jiang, Xinbai; Wu, Shijing; Shen, Jinyou; Sun, Xiuyun; Li, Jiansheng; Lu, Lude; Wang, Lianjun

    2015-01-01

    Abstract: Aerobic granules were successfully cultivated in a sequencing batch reactor (SBR), using a single bacterial strain Rhizobium sp. NJUST18 as the inoculum. NJUST18 presented as both a good pyridine degrader and an efficient autoaggregator. Stable granules with diameter of 0.5–1 mm, sludge volume index of 25.6 ± 3.6 mL g −1 and settling velocity of 37.2 ± 2.7 m h −1 , were formed in SBR following 120-day cultivation. These granules exhibited excellent pyridine degradation performance, with maximum volumetric degradation rate (V max ) varied between 1164.5 mg L −1 h −1 and 1867.4 mg L −1 h −1 . High-throughput sequencing analysis exhibited a large shift in microbial community structure, since the SBR was operated under open condition. Paracoccus and Comamonas were found to be the most predominant species in the aerobic granule system after the system had stabilized. The initially inoculated Rhizobium sp. lost its dominance during aerobic granulation. However, the inoculation of Rhizobium sp. played a key role in the start-up process of this bioaugmentation system. This study demonstrated that, in addition to the hydraulic selection pressure during settling and effluent discharge, the selection of aggregating bacterial inocula is equally important for the formation of the aerobic granule

  7. Chemo-physical evolution and microstructure features of lime treated soils

    Directory of Open Access Journals (Sweden)

    Russo Giacomo

    2016-01-01

    Full Text Available In the paper some results on the effects of chemo-physical evolution of clay-lime-water suspensions on the microstructure of a lime treated kaolin have been presented. A multi-scale investigation on the sedimentation behaviour of clay suspensions under different pore water chemistry has been developed highlighting the chemo-physical mechanisms controlling particle arrangement and the soil fabric formation. The results evidenced the key role of ionic exchange in the short term on the microstructure features of the lime treated soil.

  8. Precision neutrino oscillation physics with an intermediate baseline reactor neutrino experiment

    International Nuclear Information System (INIS)

    Choubey, Sandhya; Petcov, S.T.; Piai, M.

    2003-01-01

    We discuss the physics potential of intermediate L∼20-30 km baseline experiments at reactor facilities. Assuming that the solar neutrino oscillation parameters Δm · 2 and θ · lie in the high LMA solution region, we show that such an intermediate baseline reactor experiment can determine both Δm · 2 and θ · with a remarkably high precision. We perform also a detailed study of the sensitivity of the indicated experiment to Δm atm 2 , which drives the dominant atmospheric ν μ (ν-bar μ ) oscillations, and to θ--the neutrino mixing angle limited by the data from the CHOOZ and Palo Verde experiments. Irrespective of the actual values of Δm · 2 , we find that this experiment can improve the bounds on sin 2 θ, and, if the value of sin 2 θ is large enough, sin 2 θ > or approx. 0.02, the energy resolution of the detector is sufficiently good and if the statistics is relatively high, it can determine with extremely high precision the value of Δm atm 2 . We also explore the potential of the intermediate baseline reactor neutrino experiment for determining the type of the neutrino mass spectrum, which can be with normal or inverted hierarchy, assuming Δm · 2 to lie in the high LMA solution region. We show that the conditions under which the type of neutrino mass hierarchy can be determined are quite challenging, but are within the reach of the experiment under discussion

  9. Algal Feedback and Removal Efficiency in a Sequencing Batch Reactor Algae Process (SBAR to Treat the Antibiotic Cefradine.

    Directory of Open Access Journals (Sweden)

    Jianqiu Chen

    Full Text Available Many previous studies focused on the removal capability for contaminants when the algae grown in an unexposed, unpolluted environment and ignored whether the feedback of algae to the toxic stress influenced the removal capability in a subsequent treatment batch. The present research investigated and compared algal feedback and removal efficiency in a sequencing batch reactor algae process (SBAR to remove cefradine. Three varied pollution load conditions (10, 30 and 60 mg/L were considered. Compared with the algal characteristics in the first treatment batch at 10 and 30 mg/L, higher algal growth inhibition rates were observed in the second treatment batch (11.23% to 20.81%. In contrast, algae produced more photosynthetic pigments in response to cefradine in the second treatment batch. A better removal efficiency (76.02% was obtained during 96 h when the alga treated the antibiotic at 60 mg/L in the first treatment batch and at 30 mg/L in the second treatment batch. Additionally, the removal rate per unit algal density was also improved when the alga treated the antibiotic at 30 or 60 mg/L in the first treatment batch, respectively and at 30 mg/L in the second treatment batch. Our result indicated that the green algae were also able to adapt to varied pollution loads in different treatment batches.

  10. Algal Feedback and Removal Efficiency in a Sequencing Batch Reactor Algae Process (SBAR) to Treat the Antibiotic Cefradine

    Science.gov (United States)

    Chen, Jianqiu; Zheng, Fengzhu; Guo, Ruixin

    2015-01-01

    Many previous studies focused on the removal capability for contaminants when the algae grown in an unexposed, unpolluted environment and ignored whether the feedback of algae to the toxic stress influenced the removal capability in a subsequent treatment batch. The present research investigated and compared algal feedback and removal efficiency in a sequencing batch reactor algae process (SBAR) to remove cefradine. Three varied pollution load conditions (10, 30 and 60 mg/L) were considered. Compared with the algal characteristics in the first treatment batch at 10 and 30 mg/L, higher algal growth inhibition rates were observed in the second treatment batch (11.23% to 20.81%). In contrast, algae produced more photosynthetic pigments in response to cefradine in the second treatment batch. A better removal efficiency (76.02%) was obtained during 96 h when the alga treated the antibiotic at 60 mg/L in the first treatment batch and at 30 mg/L in the second treatment batch. Additionally, the removal rate per unit algal density was also improved when the alga treated the antibiotic at 30 or 60 mg/L in the first treatment batch, respectively and at 30 mg/L in the second treatment batch. Our result indicated that the green algae were also able to adapt to varied pollution loads in different treatment batches. PMID:26177093

  11. Parameter definition for reactor physics calculation of Obrigheim KWO PWR type reactor using the Gels and Erebus codes

    International Nuclear Information System (INIS)

    Faya, A.G.; Nakata, H.; Rodrigues, V.G.; Oosterkamp, W.J.

    1974-01-01

    The main variables for Obrigheim Reactor - KWO diffusion theory calculations, using the EREBUS code were defined. The variables under consideration were: mesh spacing for reactor description, time-step in burn-up calculation, and the temperature in both the moderator and the fuel. The best mesh spacing and time-step were defined considering the relative deviations and the computer time expended in each case. It has been verified that the error involved in the mean fuel temperature calculation (1317 0 K as given by SIEMENS and 1028 0 K as calculated by Dr. Penndorf) does not change substancially the calculation results

  12. Research on reactor physics analysis method based on Monte Carlo homogenization

    International Nuclear Information System (INIS)

    Ye Zhimin; Zhang Peng

    2014-01-01

    In order to meet the demand of nuclear energy market in the future, many new concepts of nuclear energy systems has been put forward. The traditional deterministic neutronics analysis method has been challenged in two aspects: one is the ability of generic geometry processing; the other is the multi-spectrum applicability of the multigroup cross section libraries. Due to its strong geometry modeling capability and the application of continuous energy cross section libraries, the Monte Carlo method has been widely used in reactor physics calculations, and more and more researches on Monte Carlo method has been carried out. Neutronics-thermal hydraulics coupling analysis based on Monte Carlo method has been realized. However, it still faces the problems of long computation time and slow convergence which make it not applicable to the reactor core fuel management simulations. Drawn from the deterministic core analysis method, a new two-step core analysis scheme is proposed in this work. Firstly, Monte Carlo simulations are performed for assembly, and the assembly homogenized multi-group cross sections are tallied at the same time. Secondly, the core diffusion calculations can be done with these multigroup cross sections. The new scheme can achieve high efficiency while maintain acceptable precision, so it can be used as an effective tool for the design and analysis of innovative nuclear energy systems. Numeric tests have been done in this work to verify the new scheme. (authors)

  13. Verification of Unstructured Mesh Capabilities in MCNP6 for Reactor Physics Problems

    International Nuclear Information System (INIS)

    Burke, Timothy P.; Martz, Roger L.; Kiedrowski, Brian C.; Martin, William R.

    2012-01-01

    New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructive Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.

  14. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  15. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

    2011-01-01

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  16. The implementation and evaluation of physical protection system of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Vaz, Antonio Carlos Alves

    2016-01-01

    The September 11, 2001 terrorist attacks in New York, the accident at the Fukushima nuclear power plant on March 2011 and the recent attacks in Paris on November 2015 are examples of events that justify the efforts of the International Agency of Energy Atomic - IAEA to improve security at nuclear facility. The Brazilian government has been collaborating with this project and investing resources to improve the Physical Protection System - PPS of the nuclear research reactor system, technically is associated with the elements of detection, delay and response. The PPS is an integrated system of people, equipment and procedures used to protect nuclear facilities and radioactive sources against threat, theft or sabotage. The PPS works to avoid, to mitigate or to minimize the consequences caused by these actions. This study evaluates the PPS of the reactor, identifying the vulnerabilities and suggesting ways to improve the system effectiveness. The analyses were based on the methodology developed by Sandia National Laboratories´ security experts in Albuquerque - USA, allowing the system evaluation through hypothetical and probabilistic analyzes; identifying threats, determining the targets and analyzing the possible adversaries paths. From the methodology adopted was obtained the value around 40% for PE indicator, which shows the need to improve the system to minimizing the vulnerabilities. (author)

  17. Main research results in the field of nuclear power engineering of the Nuclear Reactors and Thermal Physics Institute in 2014

    International Nuclear Information System (INIS)

    Trufanov, A.A.; Orlov, Yu.I.; Sorokin, A.P.; Chernonog, V.L.

    2015-01-01

    The main results of scientific and technological activities for last years of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in solving problems of nuclear power engineering are presented. The work have been carried out on the following problems: justification of research and development solutions and safety of NPPs with fast reactors of new generation with sodium (BN-1200, MBIR) and lead (BREST-OD-300) coolants, justification of safety of operating and advanced NPPs with WWER reactor facilities (WWER-1000, AEhS 2006, WWER-TOI), development and benchmarking of computational codes, research and development support of Beloyarsk-3 (BN-600) and Bilibino (BN-800) NPPs operation, decommissioning of AM and BR-10 research reactors, pilot scientific studies (WWER-SKD, ITER), international scientific and technical cooperation. Problems for further investigations are charted [ru

  18. Development of empirical models for performance evaluation of UASB reactors treating poultry manure wastewater under different operational conditions

    International Nuclear Information System (INIS)

    Yetilmezsoy, Kaan; Sakar, Suleyman

    2008-01-01

    A nonlinear modeling study was carried out to evaluate the performance of UASB reactors treating poultry manure wastewater under different organic and hydraulic loading conditions. Two identical pilot scale up-flow anaerobic sludge blanket (UASB) reactors (15.7 L) were run at mesophilic conditions (30-35 deg. C) in a temperature-controlled environment with three hydraulic retention times (θ) of 15.7, 12 and 8.0 days. Imposed volumetric organic loading rates (L V ) ranged from 0.65 to 4.257 kg COD/(m 3 day). The pH of the feed varied between 6.68 and 7.82. The hydraulic loading rates (L H ) were controlled between 0.105 and 0.21 m 3 /(m 2 day). The daily biogas production rates ranged between 4.2 and 29.4 L/day. High volumetric COD removal rates (R V ) ranging from 0.546 to 3.779 kg COD removed /(m 3 day) were achieved. On the basis of experimental results, two empirical models having a satisfactory correlation coefficient of about 0.9954 and 0.9416 were developed to predict daily biogas production (Q g ) and effluent COD concentration (S e ), respectively. Findings of this modeling study showed that optimal COD removals ranging from 86.3% to 90.6% were predicted with HRTs of 7.9, 9.5, 11.2, 12.6, 13.7 and 14.3 days, and L V of 1.27, 1.58, 1.78, 1.99, 2.20 and 2.45 kg COD/(m 3 day) for the corresponding influent substrate concentrations (S i ) of 10,000, 15,000, 20,000, 25,000, 30,000 and 35,000 mg/L, respectively

  19. Discrete nodal integral transport-theory method for multidimensional reactor physics and shielding calculations

    International Nuclear Information System (INIS)

    Lawrence, R.D.; Dorning, J.J.

    1980-01-01

    A coarse-mesh discrete nodal integral transport theory method has been developed for the efficient numerical solution of multidimensional transport problems of interest in reactor physics and shielding applications. The method, which is the discrete transport theory analogue and logical extension of the nodal Green's function method previously developed for multidimensional neutron diffusion problems, utilizes the same transverse integration procedure to reduce the multidimensional equations to coupled one-dimensional equations. This is followed by the conversion of the differential equations to local, one-dimensional, in-node integral equations by integrating back along neutron flight paths. One-dimensional and two-dimensional transport theory test problems have been systematically studied to verify the superior computational efficiency of the new method

  20. Physics parameter calculations for a Tandem Mirror Reactor with thermal barriers

    International Nuclear Information System (INIS)

    Boghosian, B.M.; Lappa, D.A.; Logan, B.G.

    1979-01-01

    Thermal barriers are localized reductions in potential between the plugs and the central cell, which effectively insulate trapped plug electrons from the central cell electrons. By then applying electron heating in the plug, it is possible to obtain trapped electron temperatures that are much greater than those of the central cell electrons. This, in turn, effects an increase in the plug potential and central cell confinement with a concomitant decrease in plug density and injection power. Ions trapped in the barrier by collisions are removed by the injection of neutral beams directed inside the barrier cell loss cone; these beam neutrals convert trapped barrier ions to neutrals by charge exchange permitting their escape. We describe a zero-dimensional physics model for this type of reactor, and present some preliminary results for Q

  1. Studies on solid-state physics carried out with the Saclay reactor (1962)

    International Nuclear Information System (INIS)

    Herpin, A.

    1962-01-01

    This paper deals only with solid-state physics experiments carried out on outgoing beams: rather than giving a general review of the work performed, if refers to only a few of the most important studies or those nearest completion. These are being made with the experimental beams of the two Saclay reactors EL-2, with a central flux of 10 13 n/cm 2 , and - since 1958 - EL-3, whose central flux is equal ta 10 14 n/cm 2 . The experiments are being carried out by two separate groups of physicists, employing different techniques, namely neutron diffraction using a crystal spectrometer, and inelastic scattering using a time-of-flight spectrometer. (author) [fr

  2. Organization and management of health physics support for a research reactor

    International Nuclear Information System (INIS)

    Bates, E.F.; Neff, R.D.; Randall, J.D.

    1980-01-01

    The Radiological Safety Office administers the radiological safety and surveillance programs for Texas A and M University. This program includes the assignment of a health physics group to the Texas A and M University Nuclear Science Center. By mutual agreement, the Nuclear Science Center health physics group acts as an integral part of the NSC staff which provides a system for making positive contributions to the decision-making process and the management of its time and resources to accomplish the design objectives of the radiation safety program. These personnel administer a continuous program of hazard analyses and evaluations to minimize and eliminate radiological hazards in terms of occupational exposures to radiation, contamination control, and release of radioactive effluents to the environs. This program has been effective in reducing occupational exposures to radiation in terms of total manrem expended and maintaining effluent releases to the environment at approximately 2% of the limits specified in 10CFR20. This paper presents' an organizational method for establishing an operational and functional research reactor health physics group and the resultant benefits from its contribution to the overall organization. (author)

  3. Validation and application of a physics database for fast reactor fuel cycle analysis

    International Nuclear Information System (INIS)

    McKnight, R.D.; Stillman, J.A.; Toppel, B.J.; Khalil, H.S.

    1994-01-01

    An effort has been made to automate the execution of fast reactor fuel cycle analysis, using EBR-II as a demonstration vehicle, and to validate the analysis results for application to the IFR closed fuel cycle demonstration at EBR-II and its fuel cycle facility. This effort has included: (1) the application of the standard ANL depletion codes to perform core-follow analyses for an extensive series of EBR-II runs, (2) incorporation of the EBR-II data into a physics database, (3) development and verification of software to update, maintain and verify the database files, (4) development and validation of fuel cycle models and methodology, (5) development and verification of software which utilizes this physics database to automate the application of the ANL depletion codes, methods and models to perform the core-follow analysis, and (6) validation studies of the ANL depletion codes and of their application in support of anticipated near-term operations in EBR-II and the Fuel Cycle Facility. Results of the validation tests indicate the physics database and associated analysis codes and procedures are adequate to predict required quantities in support of early phases of FCF operations

  4. Evaluation guide for the international reactor physics experiments evaluation project (IRPhEP)

    International Nuclear Information System (INIS)

    Yamaji, Akifumi

    2013-01-01

    At present, there is an urgent need to preserve integral reactor physics experimental data including separate or special effects data for nuclear energy and technology applications and the knowledge and competence contained therein. The International Reactor Physics Evaluation Project (IRPhEP) was initiated as a pilot activity in 1999 by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. While coordination and administration of the IRPhEP takes place at an international level, each participating country is responsible for the administration, technical direction, and priorities of the project within their respective countries. This document outlines the general presentation guidelines that evaluators should follow for the description of the experiments and all relevant experimental data in order to ensure the consistency between the evaluations published in the final Handbook. Publication templates will be used to ensure this consistency and will follow the general scheme below: 1 - Experiment identification number; 2- Date; 3 - Name of experiment (Purpose of experiment, Phenomena measured and scope); 4 - Name or designation of experimental programme; 5 - Description of facility; 6 - Description of test or experiment (Experimental configuration, Core life cycle, Experimental limitations or shortcomings); 7 - Phenomena measured (Description of results and analysis, Special features and characteristics of experiment, Measurement systems/methods and uncertainties); 8 - Duplicate or complementary experiments / other related experiments; 9 - Status of completion of the evaluation; 10 - References (pointer to evaluation, archive if available, otherwise generic bibliographic reference); 11 - Authors/ organisers 12 - Material available

  5. 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

  6. Aspects of Reactor Physics Research at the Victoria University of Manchester

    International Nuclear Information System (INIS)

    Harris, M.J.; Walton, D.G.

    1964-01-01

    The Nuclear Engineering Department at Manchester University was established in 1959. Since that time post-graduate reactor physics studies have gradually enlarged and developed, starting virtually from scratch; experimental studies have concentrated on light-water systems and centred on the accelerator-driven, natural-uranium, light-water exponential. The paper contains a survey of the work to date, discussion of the results obtained, outlines of proposed future work, and, as they arise in the text, descriptions of various low-cost, labour-saving experimental techniques which have been adopted. The various divisions of the work are described below. The authors have studied neutron diffusion in light water using both pulsed source and steady source methods. In the former method they have particularly stressed full harmonic analysis to the extent of actually studying the higher modes as opposed to most former work which has tried only to eliminate them. In the study of steady source methods they have concentrated on eliminating all effects from finite source and detector size, resonance activation, flux perturbation and so on. The results of both are discussed and compared. A very careful measurement of absorption cross-sections by the pulsed technique, taking care to eliminate harmonic and other effects likely to lead to error is also in progress and is described. Thermal neutron spectra in ''poisoned'' light water are being measured as a means of investigating and developing integral detector techniques. This discussion includes some interesting time- and cost-saving examples. Large foil activation and counting techniques for measuring spatially averaged neutron densities, and hence a number of reactor parameters, have been studied. Some interesting points have arisen, particularly with regard to spectrum measurement. The method makes possible many reactor physics investigations with limited resources. A low-cost natural uranium, light-water exponential has been

  7. Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Hyun, E-mail: jhkim@actbest.com [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); ACT Co., Ltd, 705 Gwanpyeong-dong, Yuseong-gu, Daejeon 305-509 (Korea, Republic of); Jin, Dong Sik [ACT Co., Ltd, 705 Gwanpyeong-dong, Yuseong-gu, Daejeon 305-509 (Korea, Republic of); Chang, Soon Heung [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1, Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of)

    2013-10-15

    Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively.

  8. Combining Voice Therapy and Physical Therapy: A Novel Approach to Treating Muscle Tension Dysphonia

    Science.gov (United States)

    Craig, Jennifer; Tomlinson, Carey; Stevens, Kristin; Kotagal, Kiran; Fornadley, Judith; Jacobson, Barbara; Garrett, C. Gaelyn; Francis, David O.

    2015-01-01

    Objective This study investigated the role of a specialized physical therapy program for muscle tension dysphonia patients as an adjunct to standard of care voice therapy. Study Design Retrospective Cohort Study Methods Adult MTD patients seen between 2007 and 2012 were identified from the clinical database. They were prescribed voice therapy and, if concomitant neck pain, adjunctive physical therapy. In a pragmatic observational cohort design, patients underwent one of four potential treatment approaches: voice therapy alone (VT), voice therapy and physical therapy (VT+PT), physical therapy alone (PT), or incomplete/no treatment. Voice handicap outcomes were compared between treatment approaches. Results Of 153 patients meeting criteria (Median age 48 years, 68% female, and 30% had fibromyalgia, chronic pain, chronic fatigue, depression, and/or anxiety), there was a similar distribution of patients with moderate or severe pre-treatment VHI scores across treatment groups (VT 45.5%, VT+PT 43.8%, PT 50%, no treatment 59.1%; p=0.45). Patients treated with VT alone had significantly greater median improvement in VHI than those not treated: 10-point vs. 2-point (p=0.02). Interestingly, median VHI improvement in patients with baseline moderate-severe VHI scores was no different between VT (10), VT+PT (8) and PT alone (10; p=0.99). Conclusions Findings show voice therapy to be an effective approach to treating MTD. Importantly, other treatment modalities incorporating physical therapy had a similar, albeit not significant, improvement in VHI. This preliminary study suggests that physical therapy techniques may have a role in the treatment of a subset of MTD patients. Larger, comparative studies are needed to better characterize the role of physical therapy in this population. PMID:26012419

  9. Proceeding of the Scientific Meeting and Presentation on Basic Research of Nuclear Science and Technology: Book I. Physics, Reactor Physics and Nuclear Instrumentation

    International Nuclear Information System (INIS)

    1996-06-01

    The proceeding contains papers presented on Scientific Meeting and Presentation on on Basic Research of Nuclear Science and Technology, held in Yogyakarta, 25-27 April 1995. This proceeding is part one from two books published for the meeting contains papers on Physics, Reactor Physics and Nuclear Instrumentation as results of research activities in National Atomic Energy Agency. There are 39 papers indexed individually. (ID)

  10. Physical model of lean suppression pressure oscillation phenomena: steam condensation in the light water reactor pressure suppression system (PSS)

    International Nuclear Information System (INIS)

    McCauley, E.W.; Holman, G.S.; Aust, E.; Schwan, H.; Vollbrandt, J.

    1980-01-01

    Using the results of large scale multivent tests conducted by GKSS, a physical model of chugging is developed. The unique combination of accurate digital data and cinematic data has provided the derivation of a detailed, quantified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena occurring during lean suppression (chugging) phases of the loss-of-coolant accident in a boiling water reactor pressure suppression system

  11. Microbial succession within an anaerobic sequencing batch biofilm reactor (ASBBR treating cane vinasse at 55ºC

    Directory of Open Access Journals (Sweden)

    Maria Magdalena Ferreira Ribas

    2009-08-01

    Full Text Available The aim of this work was to investigate the anaerobic biomass formation capable of treating vinasse from the production of sugar cane alcohol, which was evolved within an anaerobic sequencing batch biofilm reactor (ASBBR as immobilized biomass on cubes of polyurethane foam at the temperature of 55ºC. The reactor was inoculated with mesophilic granular sludge originally treating poultry slaughterhouse wastewater. The evolution of the biofilm in the polyurethane foam matrices was assessed during seven experimental phases which were thus characterized by the changes in the organic matter concentrations as COD (1.0 to 20.0 g/L. Biomass characterization proceeded with the examination of sludge samples under optical and scanning electron microscopy. The reactor showed high microbial morphological diversity along the trial. The predominance of Methanosaeta-like cells was observed up to the organic load of 2.5 gCOD/L.d. On the other hand, Methanosarcinalike microorganisms were the predominant archaeal population within the foam matrices at high organic loading ratios above 3.3 gCOD/L.d. This was suggested to be associated to a higher specific rate of acetate consumption by the later organisms.Este trabalho investigou a formação de um biofilme anaeróbio capaz de tratar vinhaça da produção de álcool de cana-de-açúcar, que evoluiu dentro de um reator operado em bateladas seqüenciais com biofilme (ASBBR tendo a biomassa imobilizada em cubos de espuma de poliuretano na temperatura de 55ºC. O reator foi inoculado com lodo granular mesofílico tratando água residuária de abatedouro de aves. A evolução do biofilme nas matrizes de espuma de poliuretano foi observada durante sete fases experimentais que foram caracterizadas por mudanças nas concentrações de matéria orgânica como DQO (1,0 a 20,0 g/L. A caracterização da biomassa foi feita por exames de amostras do lodo em microscopia ótica e eletrônica de varredura. O reator apresentou

  12. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    Science.gov (United States)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  13. Effect of florfenicol on performance and microbial community of a sequencing batch biofilm reactor treating mariculture wastewater.

    Science.gov (United States)

    Gao, Feng; Li, Zhiwei; Chang, Qingbo; Gao, Mengchun; She, Zonglian; Wu, Juan; Jin, Chunji; Zheng, Dong; Guo, Liang; Zhao, Yangguo; Wang, Sen

    2018-02-01

    The effects of florfenicol (FF) on the performance, microbial activity and microbial community of a sequencing batch biofilm reactor (SBBR) were evaluated in treating mariculture wastewater. The chemical oxygen demand (COD) and nitrogen removal were inhibited at high FF concentrations. The specific oxygen utilization rate (SOUR), specific ammonium oxidation rate (SAOR), specific nitrite oxidation rate (SNOR) and specific nitrate reduction rate (SNRR) were decreased with an increase in the FF concentration from 0 to 35 mg/L. The chemical compositions of loosely bound extracellular polymeric substances (LB-EPS) and tightly bound EPS (TB-EPS) could be affected with an increase in the FF concentration. The high-throughput sequencing indicated some obvious variations in the microbial community at different FF concentrations. The relative abundance of Nitrosomonas and Nitrospira showed a decreasing tendency with an increase in the FF concentration, suggesting that FF could affect the nitrification process of SBBR. Some genera capable of reducing nitrate to nitrogen gas could be inhibited by the addition of FF in the influent, such as Azospirillum and Hyphomicrobium.

  14. Performance evaluation of a mesophilic (37 deg. C) upflow anaerobic sludge blanket reactor in treating distiller's grains wastewater

    International Nuclear Information System (INIS)

    Gao Mengchun; She Zonglian; Jin Chunji

    2007-01-01

    The performance of a laboratory-scale upflow anaerobic sludge blanket (UASB) reactor treating distiller's grains wastewater was investigated for 420 days at 37 deg. C. After a successful start-up, 80-97.3% chemical oxygen demand (COD) removal efficiencies were achieved at hydraulic retention times (HRT) of 82-11 h with organic loading rates (OLR) of 5-48.3 kg COD m -3 d -1 . The biogas mainly consisted of methane and carbon dioxide, and the methane and carbon dioxide content in the biogas was 57-60 and 38-41%, respectively. The yield coefficient of methane production was 0.3182 l CH 4 g -1 COD removed until OLR at 33.3 kg COD m -3 d -1 , but afterwards began to decrease. The volatile fatty acid (VFA) in the effluent mainly consisted of acetate and propionate, accounting for more than 95% of total VFA as COD, and other VFA was detected at insignificant concentrations. The mesophilic granules developed in this study showed an excellent specific methanogenic activity (SMA) at 0.91 and 1.12 g methane COD g -1 VSS -1 d -1 using sucrose and acetate as individual substrates on day 200, respectively

  15. An Investigation on Cocombustion Behaviors of Hydrothermally Treated Municipal Solid Waste with Coal Using a Drop-Tube Reactor

    Directory of Open Access Journals (Sweden)

    Liang Lu

    2012-01-01

    Full Text Available This work aims at demonstrating the feasibility of replacing Indonesian coal (INC with hydrothermally treated municipal solid waste (MSWH in cocombustion with high ash Indian coal (IC. The combustion efficiencies and emissions (CO, NO of MSWH, INC and their blends with IC for a series of tests performed under a range of temperatures and air conditions were tested in a drop-tube reactor (DTR. The results showed the following. The combustion efficiency of IC was increased by blending both MSWH and INC and CO emission was reduced with increasing temperature. For NO emission, the blending of MSWH led to the increase of NO concentration whereas the effects of INC depended on the temperature. The combustion behaviors of IC-MSWH blend were comparable to those of the IC-INC blend indicating it is possible for MSWH to become a good substitute for INC supporting IC combustion. Moreover, the CO emission fell while the NO emission rose with increasing excess air for IC-MSWH blend at 900°C and the highest combustion efficiency was obtained at the excess air of 1.9. The existence of moisture in the cocombustion system of IC-MSWH blend could slightly improve the combustion efficiency, reduce CO, and increase NO.

  16. Effect of a solar Fered-Fenton system using a recirculation reactor on biologically treated landfill leachate.

    Science.gov (United States)

    Ye, Zhihong; Zhang, Hui; Yang, Lin; Wu, Luxue; Qian, Yue; Geng, Jinyao; Chen, Mengmeng

    2016-12-05

    The effects of electrochemical oxidation (EO), Fered-Fenton and solar Fered-Fenton processes using a recirculation flow system containing an electrochemical cell and a solar photo-reactor on biochemically treated landfill leachate were investigated. The most successful method was solar Fered-Fenton which achieved 66.5% COD removal after 120min treatment utilizing the optimum operating conditions of 47mM H2O2, 0.29mM Fe(2+), pH0 of 3.0 and a current density of 60mA/cm(2). The generation of hydroxyl radicals (OH) are mainly from Fered-Fenton process, which is enhanced by the introduction of renewable solar energy. Moreover, Fe(2+)/chlorine and UV/chlorine processes taking place in this system also result in additional production of OH due to the relatively high concentration of chloride ions contained in the leachate. The energy consumption was 74.5kWh/kg COD and the current efficiency was 36.4% for 2h treatment. In addition, the molecular weight (MW) distribution analysis and PARAFAC analysis of excitation emission matrix (EEM) fluorescence spectroscopy for different leachate samples indicated that the organics in the leachate were significantly degraded into either small molecular weight species or inorganics. Copyright © 2016 Elsevier B.V. All rights reserved.

  17. Individual and combined effects of organic, toxic, and hydraulic shocks on sequencing batch reactor in treating petroleum refinery wastewater.

    Science.gov (United States)

    Mizzouri, Nashwan Sh; Shaaban, Md Ghazaly

    2013-04-15

    This study analyzes the effects of toxic, hydraulic, and organic shocks on the performance of a lab-scale sequencing batch reactor (SBR) with a capacity of 5L. Petroleum refinery wastewater (PRWW) was treated with an organic loading rate (OLR) of approximately 0.3 kg chemical oxygen demand (COD)/kg MLSSd at 12.8h hydraulic retention time (HRT). A considerable variation in the COD was observed for organic, toxic, hydraulic, and combined shocks, and the worst values observed were 68.9, 77.1, 70.2, and 57.8%, respectively. Improved control of toxic shock loads of 10 and 20mg/L of chromium (VI) was identified. The system was adversely affected by the organic shock when a shock load thrice the normal value was used, and this behavior was repeated when the hydraulic shock was 4.8h HRT. The empirical recovery period was greater than the theoretical period because of the inhibitory effects of phenols, sulfides, high oil, and grease in the PRWW. The system recovery rates from the shocks were in the following order: toxic, organic, hydraulic, and combined shocks. System failure occurred when the combined shocks of organic and hydraulic were applied. The system was resumed by replacing the PRWW with glucose, and the OLR was reduced to half its initial value. Copyright © 2013 Elsevier B.V. All rights reserved.

  18. Kinetic modeling and microbial assessment by fluorescent in situ hybridization in anaerobic sequencing batch biofilm reactors treating sulfate-rich wastewater

    Directory of Open Access Journals (Sweden)

    A. J. Silva

    2011-06-01

    Full Text Available This paper reports the results of applying anaerobic sequencing batch biofilm reactors (AnSBBR for treating sulfate-rich wastewater. The reactor was filled with polyurethane foam matrices or with eucalyptus charcoal, used as the support for biomass attachment. Synthetic wastewater was prepared with two ratios between chemical oxygen demand (COD and sulfate concentration (COD/SO4(2- of 0.4 and 3.2. For a COD/SO4(2- ratio of 3.2, the AnSBBR performance was influenced by the support material used; the average levels of organic matter removal were 67% and 81% in the reactors filled with polyurethane foam and charcoal, respectively, and both support materials were associated with similar levels of sulfate reduction (above 90%. In both reactors, sulfate-reducing bacteria (SRB represented more than 65% of the bacterial community. The kinetic model indicated equilibrium between complete- and incomplete-oxidizing SRB in the reactor filled with polyurethane foam and predominantly incomplete-oxidizing SRB in the reactor filled with charcoal. Methanogenic activity seems to have been the determining factor to explain the better performance of the reactor filled with charcoal to remove organic matter at a COD/SO4(2- ratio of 3.2. For a COD/SO4(2- ratio of 0.4, low values of sulfate reduction (around 32% and low reaction rates were observed as a result of the small SRB population (about 20% of the bacterial community. Although the support material did not affect overall performance for this condition, different degradation pathways were observed; incomplete oxidation of organic matter by SRB was the main kinetic pathway and methanogenesis was negligible in both reactors.

  19. International Reactor Physics Handbook Database and Analysis Tool (IDAT) - IDAT user manual

    International Nuclear Information System (INIS)

    2013-01-01

    The IRPhEP Database and Analysis Tool (IDAT) was first released in 2013 and is included on the DVD. This database and corresponding user interface allows easy access to handbook information. Selected information from each configuration was entered into IDAT, such as the measurements performed, benchmark values, calculated values and materials specifications of the benchmark. In many cases this is supplemented with calculated data such as neutron balance data, spectra data, k-eff nuclear data sensitivities, and spatial reaction rate plots. IDAT accomplishes two main objectives: 1. Allow users to search the handbook for experimental configurations that satisfy their input criteria. 2. Allow users to trend results and identify suitable benchmarks experiments for their application. IDAT provides the user with access to several categories of calculated data, including: - 1-group neutron balance data for each configuration with individual isotope contributions in the reactor system. - Flux and other reaction rates spectra in a 299-group energy scheme. Plotting capabilities were implemented into IDAT allowing the user to compare the spectra of selected configurations in the original fine energy structure or on any user-defined broader energy structure. - Sensitivity coefficients (percent changes of k-effective due to elementary change of basic nuclear data) for the major nuclides and nuclear processes in a 238-group energy structure. IDAT is actively being developed. Those approved to access the online version of the handbook will also have access to an online version of IDAT. As May 2013 marks the first release, IDAT may contain data entry errors and omissions. The handbook remains the primary source of reactor physics benchmark data. A copy of IDAT user's manual is attached to this document. A copy of the IRPhE Handbook can be obtained on request at http://www.oecd-nea.org/science/wprs/irphe/irphe-handbook/form.html

  20. An Idea of 20% test of the Initial Core Reactor Physics

    International Nuclear Information System (INIS)

    Roh, Kyung Ho; Yang, Sung Tae; Jung, Ji Eun

    2012-01-01

    Many tests have been performed on the OPR1000 and APR1400 before commercial operation. Some of these tests were performed at reactor power levels of 20% and 50%. The CPC (Core Protection Calculator) power distribution test is one of these tests. It is performed to assure the reliability of the Core Protection Calculator System (CPCS). Through this test, SAM1 is calculated using the snapshots2. The test takes about nine hours at a reactor power level of 20% and about thirty hours at a reactor power level of 50%. SAM used at each reactor power level is as follows: 1. Reactor power of 0% ∼ 20%: Designed SAM (DSAM) 2. Reactor power of 20% ∼ 50%: SAM calculated (C-SAM) at a reactor power of 20% 3. Reactor power 50% ∼ End of Cycle : SAM calculated at a reactor power of 50% As mentioned earlier, SAM is calculated and punched into CPC to assure the reliability of CPCS. Therefore, CPC is operated having penalties with D-SAM until3 reaching a reactor power of 20%. That is, the penalty of CPC will be removed when SAM is calculated and punched into the CPC at a reactor power of 20%. But these penalties are considered to be removed after a reactor power of 50% test in order to maintain the conservatism of the CPC. This is done because the final values calculated using C-SAM, in contrast to those calculated using SAM, a reactor power of 50%, are not correct. This paper began from an idea, 'If so, what would happen if we removed the CPC power distribution test at a reactor power of 20%?'

  1. Present status of reactor physics in the United States and Japan-II. 1. Deterministic Transport Methods for Reactor Analysis

    International Nuclear Information System (INIS)

    Adams, Marvin L.

    2001-01-01

    We discuss deterministic transport methods used today in neutronic analysis of nuclear reactors. This discussion is not exhaustive; our goal is to provide an overview of the methods that are most widely used for analyzing light water reactors (LWRs) and that (in our opinion) hold the most promise for the future. The current practice of LWR analysis involves the following steps: 1. Evaluate cross sections from measurements and models. 2. Obtain weighted-average cross sections over dozens to hundreds of energy intervals; the result is a 'fine-group' cross-section set. 3. [Optional] Modify the fine-group set: Further collapse it using information specific to your class of reactors and/or alter parameters so that computations better agree with experiments. The result is a 'many-group library'. 4. Perform pin cell transport calculations (usually one-dimensional cylindrical); use the results to collapse the many-group library to a medium-group set, and/or spatially average the cross sections over the pin cells. 5. Perform assembly-level transport calculations with the medium-group set. It is becoming common practice to use essentially exact geometry (no pin cell homogenization). It may soon become common to skip step 4 and use the many-group library. The output is a library of few-group cross sections, spatially averaged over the assembly, parameterized to cover the full range of operating conditions. 6. Perform full-core calculations with few-group diffusion theory that contains significant homogenizations and limited transport corrections. We discuss steps 4, 5, and 6 and focus mainly on step 5. One cannot review a large topic in a short summary without simplifying reality, omitting important details, and neglecting some methods that deserve attention; for this we apologize in advance. (author)

  2. Optimization of micro-aeration intensity in acidogenic reactor of a two-phase anaerobic digester treating food waste

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Suyun [Department of Environmental and Low-Carbon Science, School of Environment and Architecture, University of Shanghai for Science and Technology, Shanghai (China); Sino-Forest Applied Research Centre for Pearl River Delta Environment, Department of Biology, Hong Kong Baptist University, Hong Kong Special Administrative Region (Hong Kong); Selvam, Ammaiyappan [Sino-Forest Applied Research Centre for Pearl River Delta Environment, Department of Biology, Hong Kong Baptist University, Hong Kong Special Administrative Region (Hong Kong); Wong, Jonathan W.C., E-mail: jwcwong@hkbu.edu.hk [Sino-Forest Applied Research Centre for Pearl River Delta Environment, Department of Biology, Hong Kong Baptist University, Hong Kong Special Administrative Region (Hong Kong)

    2014-02-15

    Highlights: • Effect of micro-aeration on acidogenesis and hydrolysis of food waste was investigated. • Micro-aeration at 258 L-air/kg TS/d increased the VFAs production 3-fold. • High aeration leads to loss of substrate through microbial biomass and respiration. • Optimum aeration increased methane recovery while high aeration intensity reduced methane yield. - Abstract: Micro-aeration is known to promote the activities of hydrolytic exo-enzymes and used as a strategy to improve the hydrolysis of particulate substrate. The effect of different micro-aeration rates, 0, 129, 258, and 387 L-air/kg TS/d (denoted as LBR-AN, LBR-6h, LBR-3h and LBR-2h, respectively) on the solubilization of food waste was evaluated at 35 °C in four leach bed reactors (LBR) coupled with methanogenic upflow anaerobic sludge blanket (UASB) reactor. Results indicate that the intensity of micro-aeration influenced the hydrolysis and methane yield. Adequate micro-aeration intensity in LBR-3h and LBR-2h significantly enhanced the carbohydrate and protein hydrolysis by 21–27% and 38–64% respectively. Due to the accelerated acidogenesis, more than 3-fold of acetic acid and butyric acid were produced in LBR-3h as compared to the anaerobic treatment LBR-AN resulting in the maximum methane yield of 0.27 L CH{sub 4}/g VS{sub added} in the UASB. The performance of LBR-6h with inadequate aeration was similar to that of LBR-AN with a comparable hydrolysis degree. Nevertheless, higher aeration intensity in LBR-2h was also unfavorable for methane yield due to significant biomass generation and CO{sub 2} respiration of up to 18.5% and 32.8% of the total soluble hydrolysate, respectively. To conclude, appropriate micro-aeration rate can promote the hydrolysis of solid organic waste and methane yield without undesirable carbon loss and an aeration intensity of 258 L-air/kg TS/d is recommended for acidogenic LBR treating food waste.

  3. Development of empirical models for performance evaluation of UASB reactors treating poultry manure wastewater under different operational conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yetilmezsoy, Kaan [Department of Environmental Engineering, Yildiz Technical University, 34349 Yildiz, Besiktas, Istanbul (Turkey)], E-mail: yetilmez@yildiz.edu.tr; Sakar, Suleyman [Department of Environmental Engineering, Yildiz Technical University, 34349 Yildiz, Besiktas, Istanbul (Turkey)

    2008-05-01

    A nonlinear modeling study was carried out to evaluate the performance of UASB reactors treating poultry manure wastewater under different organic and hydraulic loading conditions. Two identical pilot scale up-flow anaerobic sludge blanket (UASB) reactors (15.7 L) were run at mesophilic conditions (30-35 deg. C) in a temperature-controlled environment with three hydraulic retention times ({theta}) of 15.7, 12 and 8.0 days. Imposed volumetric organic loading rates (L{sub V}) ranged from 0.65 to 4.257 kg COD/(m{sup 3} day). The pH of the feed varied between 6.68 and 7.82. The hydraulic loading rates (L{sub H}) were controlled between 0.105 and 0.21 m{sup 3}/(m{sup 2} day). The daily biogas production rates ranged between 4.2 and 29.4 L/day. High volumetric COD removal rates (R{sub V}) ranging from 0.546 to 3.779 kg COD{sub removed}/(m{sup 3} day) were achieved. On the basis of experimental results, two empirical models having a satisfactory correlation coefficient of about 0.9954 and 0.9416 were developed to predict daily biogas production (Q{sub g}) and effluent COD concentration (S{sub e}), respectively. Findings of this modeling study showed that optimal COD removals ranging from 86.3% to 90.6% were predicted with HRTs of 7.9, 9.5, 11.2, 12.6, 13.7 and 14.3 days, and L{sub V} of 1.27, 1.58, 1.78, 1.99, 2.20 and 2.45 kg COD/(m{sup 3} day) for the corresponding influent substrate concentrations (S{sub i}) of 10,000, 15,000, 20,000, 25,000, 30,000 and 35,000 mg/L, respectively.

  4. Optimization of micro-aeration intensity in acidogenic reactor of a two-phase anaerobic digester treating food waste

    International Nuclear Information System (INIS)

    Xu, Suyun; Selvam, Ammaiyappan; Wong, Jonathan W.C.

    2014-01-01

    Highlights: • Effect of micro-aeration on acidogenesis and hydrolysis of food waste was investigated. • Micro-aeration at 258 L-air/kg TS/d increased the VFAs production 3-fold. • High aeration leads to loss of substrate through microbial biomass and respiration. • Optimum aeration increased methane recovery while high aeration intensity reduced methane yield. - Abstract: Micro-aeration is known to promote the activities of hydrolytic exo-enzymes and used as a strategy to improve the hydrolysis of particulate substrate. The effect of different micro-aeration rates, 0, 129, 258, and 387 L-air/kg TS/d (denoted as LBR-AN, LBR-6h, LBR-3h and LBR-2h, respectively) on the solubilization of food waste was evaluated at 35 °C in four leach bed reactors (LBR) coupled with methanogenic upflow anaerobic sludge blanket (UASB) reactor. Results indicate that the intensity of micro-aeration influenced the hydrolysis and methane yield. Adequate micro-aeration intensity in LBR-3h and LBR-2h significantly enhanced the carbohydrate and protein hydrolysis by 21–27% and 38–64% respectively. Due to the accelerated acidogenesis, more than 3-fold of acetic acid and butyric acid were produced in LBR-3h as compared to the anaerobic treatment LBR-AN resulting in the maximum methane yield of 0.27 L CH 4 /g VS added in the UASB. The performance of LBR-6h with inadequate aeration was similar to that of LBR-AN with a comparable hydrolysis degree. Nevertheless, higher aeration intensity in LBR-2h was also unfavorable for methane yield due to significant biomass generation and CO 2 respiration of up to 18.5% and 32.8% of the total soluble hydrolysate, respectively. To conclude, appropriate micro-aeration rate can promote the hydrolysis of solid organic waste and methane yield without undesirable carbon loss and an aeration intensity of 258 L-air/kg TS/d is recommended for acidogenic LBR treating food waste

  5. The development of the nuclear physics in Latvia II. The building of the Research Nuclear Reactor IRT

    International Nuclear Information System (INIS)

    Ulmanis, U.

    2004-01-01

    Nuclear research reactor IRT of the Academy of Sciences was built near Riga in Salaspils. IRT is pool aqueous - aqueous reactor with nuclear fuel U-235 contained elements, located in the core at a depth of ∼ 7 m under distilled water. Ten horizontal and 10-15 vertical experimental channels are employed in experimental research with the use of neutron fluxes. For the research with gamma rays is constructed radiation loop facility with liquid In-Ga-SN solid solution as intensive gamma-ray sources. Main activities of IRT are to conduct research in nuclear spectroscopy, neutron activation analysis, neutron diffraction and radiation physics, chemistry and biology. (authors)

  6. Landslides as weathering reactors; links between physical erosion and weathering in rapidly eroding mountain belts

    Science.gov (United States)

    Emberson, R.; Hovius, N.; Galy, A.

    2014-12-01

    The link between physical erosion and chemical weathering is generally modelled with a surface-blanketing weathering zone, where the supply of fresh minerals is tied to the average rate of denudation. In very fast eroding environments, however, sediment production is dominated by landsliding, which acts in a stochastic fashion across the landscape, contrasting strongly with more uniform denudation models. If physical erosion is a driver of weathering at the highest erosion rates, then an alternative weathering model is required. Here we show that landslides can be effective 'weathering reactors'. Previous work modelling the effect of landslides on chemical weathering (Gabet 2007) considered the fresh bedrock surfaces exposed in landslide scars. However, fracturing during the landslide motion generates fresh surfaces, the total surface area of which exceeds that of the exposed scar by many orders of magnitude. Moreover, landslides introduce concavity into hillslopes, which acts to catch precipitation. This is funnelled into a deposit of highly fragmented rock mass with large reactive surface area and limited hydraulic conductivity (Lo et al. 2007). This allows percolating water reaction time for chemical weathering; any admixture of macerated organic debris could yield organic acid to further accelerate weathering. In the South island of New Zealand, seepage from recent landslide deposits has systematically high solute concentrations, far outstripping concentration in runoff from locations where soils are present. River total dissolved load in the western Southern Alps is highly correlated with the rate of recent (<35yrs) landsliding, suggesting that landslides are the dominant locus of weathering in this rapidly eroding landscape. A tight link between landsliding and weathering implies that localized weathering migrates through the landscape with physical erosion; this contrasts with persistent and ubiquitous weathering associated with soil production. Solute

  7. Tumor treating fields therapy device for glioblastoma: physics and clinical practice considerations.

    Science.gov (United States)

    Lok, Edwin; Swanson, Kenneth D; Wong, Eric T

    2015-01-01

    Alternating electric fields therapy, as delivered by the tumor treating fields device, is a new modality of cancer treatment that has been approved by the US FDA for recurrent glioblastoma. At a frequency of 200 kHz, these fields emanate from transducer arrays on the surface of the patient's scalp into the brain and perturb processes necessary for cytokinesis during tumor cell mitosis. In the registration Phase III trial for recurrent glioblastoma patients, the efficacy of the tumor treating fields as monotherapy was equivalent to chemotherapy, while scalp irritation was its major adverse event compared with systemic toxicities that were associated with cytotoxic chemotherapies. Alternating electric fields therapy is, therefore, an essential option for the treatment of recurrent glioblastoma. Here, we summarize our current knowledge of the physics, cell biology and clinical data supporting the use of the tumor treating fields therapy.

  8. Morphological study of biomass during the start-up period of a fixed-bed anaerobic reactor treating domestic sewage

    Directory of Open Access Journals (Sweden)

    Cláudio Antonio Andrade Lima

    2005-09-01

    Full Text Available This work focused on a morphological study of the microorganisms attached to polyurethane foam matrices in a horizontal-flow anaerobic immobilized biomass (HAIB reactor treating domestic sewage. The experiments consisted of monitoring the biomass colonization process of foam matrices in terms of the amount of retained biomass and the morphological characteristics of the cells attached to the support during the start-up period. Non-fluorescent rods and cocci were found to predominate in the process of attachment to the polyurethane foam surface. From the 10th week of operation onwards, an increase was observed in the morphological diversity, mainly due to rods, cocci, and Methanosaeta-like archaeal cells. Hydrodynamic problems, such as bed clogging and channeling occurred in the fixed-bed reactor, mainly due to the production of extracellular polymeric substances and their accumulation in the interstices of the bed causing a gradual deterioration of its performance, which eventually led to the system's collapse. These results demonstrated the importance and usefulness of monitoring the dynamics of the formation of biofilm during the start-up period of HAIB reactors, since it allowed the identification of operational problems.Este trabalho apresenta um estudo morfológico de microrganismos aderidos à espuma de poliuretano em reator anaeróbio horizontal de leito fixo (RAHLF, aplicado ao tratamento de esgoto sanitário. O processo de colonização do suporte pela biomassa anaeróbia e as características morfológicas das células aderidas foram monitorados durante o período de partida do reator. Bacilos e cocos não fluorescentes foram predominantes no processo de aderência direta à espuma de poliuretano. Aumento na diversidade biológica foi observado a partir da 10ª semana de operação do reator, com predominância de bacilos, cocos e arqueas metanogênicas semelhantes a Methanosaeta. Problemas hidrodinâmicos, tais como formação de

  9. Effect of extracellular enzyme activity on digestion performance of mesophilic UASB reactor treating high-strength municipal wastewater.

    Science.gov (United States)

    Turkdogan-Aydinol, F Ilter; Yetilmezsoy, Kaan; Comez, Sezen

    2011-05-01

    Effect of extracellular enzyme activity on digestion performance of up-flow anaerobic sludge blanket (UASB) reactor was investigated for enhancement of anaerobic treatability of municipal wastewater. Two identical UASB reactors (9 L), namely Reactor-A (without enzyme addition) and Reactor-B (with enzyme addition),were simultaneously operated at mesophilic conditions (32 ± 2°C) with a hydraulic retention time of 24 h. Preliminary test results showed that the highest total chemical oxygen demand (TCOD) removal were achieved with an extracellular enzyme dosage of 0.2 mL/L. In the activation period of the extracellular enzyme (on days 186-212), while Reactor-A removed up to 69.3% of TCOD and 55.9% of soluble chemical oxygen demand (SCOD), Reactor-B effectively removed up to 81.9% of TCOD and 72.2% of SCOD. The average VFA/alkalinity ratios were determined to be about 0.40 (±0.03) and 0.28 (±0.08) for Reactor-A and Reactor-B, respectively.

  10. Responses of the biogas process to pulses of oleate in reactors treating mixtures of cattle and pig manure

    DEFF Research Database (Denmark)

    Nielsen, Henrik Bjørn; Ahring, Birgitte Kiær

    2006-01-01

    of oleate. Following pulses of 0.5 and 1.0 g oleate/L, the most distinct increase in volatile fatty acid (VFA) concentrations were observed in the reactor with the lowest TS/VS content. This suggests a higher adsorption of oleate on the surfaces of biofibers in the reactor with the highest TS/VS and a less...

  11. Effect of tryptone and ammonia on the biogas process in continuously stirred tank reactors treating cattle manure

    DEFF Research Database (Denmark)

    Nielsen, Hanne Bjerg; Ahring, Birgitte Kiær

    2007-01-01

    Two themophilic continuously stirred tank reactors, R1 and Two thermophilic continuously stirred tank reactors, R1 and R2, were subject to pulses of tryptone and ammonia. R1 was operated at an ammonia-N concentration of 3.0 g l(-1) and R2 was operated at an ammonia-N concentration of 1.7 g l(-1)....

  12. Physics study on recycling of ThO2/UO2 fuel in CANDU reactors through dry reprocess technology

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Park, Chang Je; Jeong, Chang Joon

    2003-06-01

    The dry process fuel technology has high proliferation-resistance which is one of important goals of Generation-IV (Gen-IV) reactor development. It is expected that the dry process fuel technology can be applied not only to existing nuclear systems but also to future nuclear systems. In this report, the homogeneous ThO 2 -UO 2 fuel cycle option of a CANDU reactor has been studied, including the physics analysis of recycling spent fuel. Reactivity swing and variation of isotopic content with irradiation are reported for various cases of initial uranium loadings. It was found that natural uranium saving increases significantly by recycling thorium/uranium fuel and it is feasible to recycle thorium with the dry process technology in a CANDU reactor. It is, however, required to further investigate the dry process that can be applied to the thorium-abundant dioxide fuel

  13. Treating the Initial Physical Reflex of Misophonia With the Neural Repatterning Technique: A Counterconditioning Procedure

    Directory of Open Access Journals (Sweden)

    Thomas H. Dozier

    2015-10-01

    Full Text Available Misophonia is a condition in which a person has an acute emotional response of anger or disgust to a commonly occurring innocuous auditory or visual stimulus referred to as a trigger. This case details the effective treatment of misophonia in a young woman that included a counterconditioning treatment called the Neural Repatterning Technique (NRT, which combines a continuous positive stimulus and a reduced intensity, intermittent trigger. The treatment was delivered via the Misophonia Trigger Tamer smartphone app and all treatments were conducted independently by the patient. In this patient, the trigger elicited a physical reflex of contraction of the flexor digitorum profundus, which caused her to clench her fist. To enhance the effect of the NRT treatment, Progressive Muscle Relaxation was incorporated to increase her ability to deliberately relax the affected muscle during treatment. During NRT treatment sessions, the patient experienced a weak physical reflex to the reduced trigger stimulus but no emotional response. Her emotional response of misophonia was not treated, but when the physical reflex extinguished, the emotional response also extinguished. This case indicates that the misophonic response includes a Pavlovian-conditioned physical reflex. It is proposed that the trigger elicited the physical reflex and the physical reflex then elicited the conditioned emotional response that is characteristic of misophonia. Because of the conditioned reflex nature of misophonia, it is proposed that a more appropriate name for this disorder would be Conditioned Aversive Reflex Disorder.

  14. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  15. A study of the literature on nodal methods in reactor physics calculations

    International Nuclear Information System (INIS)

    Van de Wetering, T.F.H.

    1993-01-01

    During the last few decades several calculation methods have been developed for the three-dimensional analysis of a reactor core. A literature survey was carried out to gain insights in the starting points and method of operation of the advanced nodal methods. These methods are applied in reactor core analyses of large nuclear power reactors, because of their high computing speed. The so-called Nodal-Expansion method is described in detail

  16. Biotic and abiotic processes contribute to successful anaerobic degradation of cyanide by UASB reactor biomass treating brewery waste water.

    Science.gov (United States)

    Novak, Domen; Franke-Whittle, Ingrid H; Pirc, Elizabeta Tratar; Jerman, Vesna; Insam, Heribert; Logar, Romana Marinšek; Stres, Blaž

    2013-07-01

    In contrast to the general aerobic detoxification of industrial effluents containing cyanide, anaerobic cyanide degradation is not well understood, including the microbial communities involved. To address this knowledge gap, this study measured anaerobic cyanide degradation and the rearrangements in bacterial and archaeal microbial communities in an upflow anaerobic sludge blanket (UASB) reactor biomass treating brewery waste water using bio-methane potential assays, molecular profiling, sequencing and microarray approaches. Successful biogas formation and cyanide removal without inhibition were observed at cyanide concentrations up to 5 mg l(-1). At 8.5 mg l(-1) cyanide, there was a 22 day lag phase in microbial activity, but subsequent methane production rates were equivalent to when 5 mg l(-1) was used. The higher cumulative methane production in cyanide-amended samples indicated that part of the biogas was derived from cyanide degradation. Anaerobic degradation of cyanide using autoclaved UASB biomass proceeded at a rate more than two times lower than when UASB biomass was not autoclaved, indicating that anaerobic cyanide degradation was in fact a combination of simultaneous abiotic and biotic processes. Phylogenetic analyses of bacterial and archaeal 16S rRNA genes for the first time identified and linked the bacterial phylum Firmicutes and the archaeal genus Methanosarcina sp. as important microbial groups involved in cyanide degradation. Methanogenic activity of unadapted granulated biomass was detected at higher cyanide concentrations than reported previously for the unadapted suspended biomass, making the aggregated structure and predominantly hydrogenotrophic nature of methanogenic community important features in cyanide degradation. The combination of brewery waste water and cyanide substrate was thus shown to be of high interest for industrial level anaerobic cyanide degradation. Copyright © 2013 Elsevier Ltd. All rights reserved.

  17. Individual and combined effects of organic, toxic, and hydraulic shocks on sequencing batch reactor in treating petroleum refinery wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Mizzouri, Nashwan Sh., E-mail: nashwan_mizzouri@yahoo.com [Department of Civil Engineering, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia); Department of Civil Engineering, University of Duhok, Kurdistan (Iraq); Shaaban, Md Ghazaly [Department of Civil Engineering, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia)

    2013-04-15

    Highlights: ► This research focuses on the combined impact of shock loads on the PRWW treatment. ► System failure resulted when combined shock of organic and hydraulic was applied. ► Recovery was achieved by replacing glucose with PRWW and OLR was decreased to half. ► Worst COD removals were 68.9, and 57.8% for organic, and combined shocks. -- Abstract: This study analyzes the effects of toxic, hydraulic, and organic shocks on the performance of a lab-scale sequencing batch reactor (SBR) with a capacity of 5 L. Petroleum refinery wastewater (PRWW) was treated with an organic loading rate (OLR) of approximately 0.3 kg chemical oxygen demand (COD)/kg MLSS d at 12.8 h hydraulic retention time (HRT). A considerable variation in the COD was observed for organic, toxic, hydraulic, and combined shocks, and the worst values observed were 68.9, 77.1, 70.2, and 57.8%, respectively. Improved control of toxic shock loads of 10 and 20 mg/L of chromium (VI) was identified. The system was adversely affected by the organic shock when a shock load thrice the normal value was used, and this behavior was repeated when the hydraulic shock was 4.8 h HRT. The empirical recovery period was greater than the theoretical period because of the inhibitory effects of phenols, sulfides, high oil, and grease in the PRWW. The system recovery rates from the shocks were in the following order: toxic, organic, hydraulic, and combined shocks. System failure occurred when the combined shocks of organic and hydraulic were applied. The system was resumed by replacing the PRWW with glucose, and the OLR was reduced to half its initial value.

  18. Application of magnetic OMS-2 in sequencing batch reactor for treating dye wastewater as a modulator of microbial community.

    Science.gov (United States)

    Pan, Fei; Yu, Yang; Xu, Aihua; Xia, Dongsheng; Sun, Youmin; Cai, Zhengqing; Liu, Wen; Fu, Jie

    2017-10-15

    The potential and mechanism of synthesized magnetic octahedral molecular sieve (Fe 3 O 4 @OMS-2) nanoparticles in enhancing the aerobic microbial ability of sequencing batch reactor (SBR) for treating dye wastewater have been revealed in this study. The addition of Fe 3 O 4 @OMS-2 of 0.25g/L enhanced the decolorization of SBRs with an operation cycle of 24h by more than 20%. The 16S rRNA gene high-throughput sequencing indicated Fe 3 O 4 @OMS-2 increased the microbial richness and diversity of SBRs, and more importantly, promoted the potential dye-degrading bacteria. After a series of enriching and screening, four bacterial strains with the considerable decolorizing ability were isolated from SBRs, designating Alcaligenes faecalis FP-G1, Bacillus aryabhattai FP-F1, Escherichia fergusonii FP-D1 and Rhodococcus ruber FP-E1, respectively. The growth and decolorization of these pure strains were promoted in the presence of Fe 3 O 4 @OMS-2, which agrees with the result of high-throughput sequencing. Monitoring dissolved Fe/Mn ions and investigating the change of oxidation states of Fe/Mn species discovered OMS-2 composition played the critical role in modulating the microbial community. The significant enhancement of Mn-oxidizing/-reducing bacteria suggested microbial Mn redox may be the key action mechanism of Fe 3 O 4 @OMS-2, which can provide numerous benefits for the microbial community and decolorization of SBRs. Copyright © 2017 Elsevier B.V. All rights reserved.

  19. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-15

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  20. Evaluation of the physical protection system of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Vaz, Antonio C.A.; Conti, Thadeu das N.

    2013-01-01

    The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

  1. Training at the masters degree level in physics and technology of nuclear reactors in the uk

    International Nuclear Information System (INIS)

    Weaver, D.R.

    2000-01-01

    This paper discusses the current situation of university-based training for the nuclear power industry in the UK, drawing on information gathered as part of the survey for a review currently being undertaken by the Committee for Technical and Economic Studies on Nuclear Energy Development and Fuel Cycle (NDC) of the Nuclear Energy Agency (NEA) of the OECD. A particular focus will be the Physics and Technology of Nuclear Reactors MSc course at the University of Birmingham. In the past there were other similar MSc courses in the UK, but through the evolution of time the Birmingham course is now unique in its role of providing masters level training so specifically aimed at the commercial nuclear programme. Mention will, however, be made of other training at the postgraduate level elsewhere in the UK. A description is given of the need to consider a new form of relationship between industry and university in order to provide optimise the provision of masters level training. (author)

  2. ARES: A Parallel Discrete Ordinates Transport Code for Radiation Shielding Applications and Reactor Physics Analysis

    Directory of Open Access Journals (Sweden)

    Yixue Chen

    2017-01-01

    Full Text Available ARES is a multidimensional parallel discrete ordinates particle transport code with arbitrary order anisotropic scattering. It can be applied to a wide variety of radiation shielding calculations and reactor physics analysis. ARES uses state-of-the-art solution methods to obtain accurate solutions to the linear Boltzmann transport equation. A multigroup discretization is applied in energy. The code allows multiple spatial discretization schemes and solution methodologies. ARES currently provides diamond difference with or without linear-zero flux fixup, theta weighted, directional theta weighted, exponential directional weighted, and linear discontinuous finite element spatial differencing schemes. Discrete ordinates differencing in angle and spherical harmonics expansion of the scattering source are adopted. First collision source method is used to eliminate or mitigate the ray effects. Traditional source iteration and Krylov iterative method preconditioned with diffusion synthetic acceleration are applied to solve the linear system of equations. ARES uses the Koch-Baker-Alcouffe parallel sweep algorithm to obtain high parallel efficiency. Verification and validation for the ARES transport code system have been done by lots of benchmarks. In this paper, ARES solutions to the HBR-2 benchmark and C5G7 benchmarks are in excellent agreement with published results. Numerical results are presented which demonstrate the accuracy and efficiency of these methods.

  3. New approach to invariant-embedding methods in reactor physics calculations

    International Nuclear Information System (INIS)

    Forsbacka, M.J.; Rydin, R.A.

    1997-01-01

    Invariant-embedding methods offer an alternative approach to modeling physical phenomena and solving mathematical problems. Invariant embedding allows one to express traditional boundary-value problems as initial-value problems. In doing this, one effectively reformulates a problem to be solved in terms of an embedding parameter. In this paper, a hybrid method that consists of Monte Carlo-generated response functions that describe the neutronic properties of local spatial cells are coupled together in a global reactor model using the invariant embedding methodology, where the system multiplication factor k eff is used as the embedding parameter. Thus, k eff is computed directly rather than as the result of a secondary eigenvalue calculation. Because the response functions can represent any arbitrary material distribution within a local cell, this method shows promise to accurately assess the change in reactivity due to core disruptive accidents and other changes in system configuration such as changing control rod positions. This paper reports a series of proof-of-concept calculations that assess this method

  4. Armouring facility? Nuclear-weapon and reactor reseach at the Kaiser-Wilhelm Institute for Physics

    International Nuclear Information System (INIS)

    Hachtmann, R.; Walker, M.

    2005-01-01

    The Kaiser Wilhelm Institute for Physics is best known as the place where Werner Heisenberg worked on nuclear weapons for Hitler. Although this is essentially true, there is more to the story. At the start of World War II this institute was taken over by the German Army Ordnance to be the central, but not exclusive site for a research project into the economic and military applications of nuclear fission. The Army physicist Kurt Diebner was installed in the institute as its commissarial director. Heisenberg was affiliated with the institute as an advisor at first, and became the director in 1942. Heisenberg and his colleagues, including in particular Karl-Heinz Hoecker, Carl Friedrich von Weizsaecker, and Karl Wirtz, worked on nuclear reactors and isotope separation with the clear knowledge that these were two different paths to atomic bombs [Atombomben]. However, they were clearly ambivalent about what they were doing. New documents recently returned from Russian archives shed new light on this work and the scientists' motivations. (orig.)

  5. Physical protection of shipments of irradiated reactor fuel; Interim guidance. Regulatory report

    International Nuclear Information System (INIS)

    1980-06-01

    During May, 1979, the U.S. Nuclear Regulatory Commission approved for issuance in effective form new interim regulations for strengthening the protection of spent fuel shipments against sabotage and diversion. The new regulations were issued without benefit of public comment, but comments from the public were solicited after the effective date. Based upon the public comments received, the interim regulations were amended and reissued in effective form as a final interim rule in May, 1980. The present document supersedes a previously issued interim guidance document, NUREG-0561 (June, 1979) which accompanied the original rule. This report has been revised to conform to the new interim regulations on the physical protection of shipments of irradiated reactor fuel which are likely to remain in effect until the completion of an ongoing research program concerning the response of spent fuel to certain forms of sabotage, at which time the regulations may be rescinded, modified or made permanent, as appropriate. This report discusses the amended regulations and provides a basis on which licensees can develop an acceptable interim program for the protection of spent fuel shipments

  6. The use of active learning strategies in the instruction of Reactor Physics concepts

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Michael A.

    2000-01-01

    Each of the Active Learning strategies employed to teach Reactor Physics material has been or promises to be instructionally successful. The Cooperative Group strategy has demonstrated a statistically significant increase in student performance on the unit exam in teaching conceptually difficult, transport and diffusion theory material. However, this result was achieved at the expense of a modest increase in class time. The Tutorial CBI programs have enabled learning equally as well as classroom lectures without the direct intervention of an instructor. Thus, the Tutorials have been successful as homework assignments, releasing classroom time for other instruction. However, the time required for development of these tools was large, on the order of two hundred hours per hour of instruction. The initial introduction of the Case-Based strategy was roughly as effective as the traditional classroom instruction. Case-Based learning could well, after important modifications, perform better than traditional instruction. A larger percentage of the students prefer active learning strategies than prefer traditional lecture presentations. Student preferences for the active strategies were particularly strong when they believed that the strategies helped them learn the material better than they would have by using a lecture format. In some cases, students also preferred the active strategies because they were different from traditional instruction, a change of pace. Some students preferred lectures to CBI instruction, primarily because the CBI did not afford them the opportunity to question the instructor during the presentation.

  7. The use of active learning strategies in the instruction of Reactor Physics concepts

    International Nuclear Information System (INIS)

    Robinson, Michael A.

    2000-01-01

    Each of the Active Learning strategies employed to teach Reactor Physics material has been or promises to be instructionally successful. The Cooperative Group strategy has demonstrated a statistically significant increase in student performance on the unit exam in teaching conceptually difficult, transport and diffusion theory material. However, this result was achieved at the expense of a modest increase in class time. The Tutorial CBI programs have enabled learning equally as well as classroom lectures without the direct intervention of an instructor. Thus, the Tutorials have been successful as homework assignments, releasing classroom time for other instruction. However, the time required for development of these tools was large, on the order of two hundred hours per hour of instruction. The initial introduction of the Case-Based strategy was roughly as effective as the traditional classroom instruction. Case-Based learning could well, after important modifications, perform better than traditional instruction. A larger percentage of the students prefer active learning strategies than prefer traditional lecture presentations. Student preferences for the active strategies were particularly strong when they believed that the strategies helped them learn the material better than they would have by using a lecture format. In some cases, students also preferred the active strategies because they were different from traditional instruction, a change of pace. Some students preferred lectures to CBI instruction, primarily because the CBI did not afford them the opportunity to question the instructor during the presentation

  8. Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kontogeorgakos, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Derstine, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Bauer, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-06-01

    The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO2 particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.

  9. A Special Topic From Nuclear Reactor Dynamics for the Undergraduate Physics Curriculum

    Science.gov (United States)

    Sevenich, R. A.

    1977-01-01

    Presents an intuitive derivation of the point reactor equations followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Suggests several computer simulations involving the effect of control rod motion on reactor power. (MLH)

  10. The physical and chemical properties of plasma treated ultra-high-molecular-weight polyethylene fibers

    DEFF Research Database (Denmark)

    Kusano, Yukihiro; Teodoru, Steluta; Hansen, Charles M.

    2011-01-01

    polymer assures maximum physical adhesion to transfer loads uniformly. Plasma treatment of ultra-high-molecular-weight polyethylene (UHMWPE) fibers is shown to significantly increase the amount of oxygen in the surface. There are two distinct types of surfaces in both the plasma treated and the untreated......A uniform and smooth transfer of stresses across the polymer matrix/fiber interface is enhanced when adhesion between the matrix and fiber surface is optimized. In the absence of covalent bonds matching the Hansen solubility (cohesion) parameters (HSP) of the fiber surface with the HSP of a matrix...... UHMWPE fibers. One type is typical of polyethylene (PE) polymers while the other is characteristic of the oxygenated surface at much higher values of HSP. The oxygenated surface of the plasma treated fibers has the HSP δD, δP, and δH equal to 16.5, 15.3, and 8.2, compared to the pure PE surface with HSP...

  11. Physical modelling of the composting environment: A review. Part 1: Reactor systems

    International Nuclear Information System (INIS)

    Mason, I.G.; Milke, M.W.

    2005-01-01

    In this paper, laboratory- and pilot-scale reactors used for investigation of the composting process are described and their characteristics and application reviewed. Reactor types were categorised by the present authors as fixed-temperature, self-heating, controlled temperature difference and controlled heat flux, depending upon the means of management of heat flux through vessel walls. The review indicated that fixed-temperature reactors have significant applications in studying reaction rates and other phenomena, but may self-heat to higher temperatures during the process. Self-heating laboratory-scale reactors, although inexpensive and uncomplicated, were shown to typically suffer from disproportionately large losses through the walls, even with substantial insulation present. At pilot scale, however, even moderately insulated self-heating reactors are able to reproduce wall losses similar to those reported for full-scale systems, and a simple technique for estimation of insulation requirements for self-heating reactors is presented. In contrast, controlled temperature difference and controlled heat flux laboratory reactors can provide spatial temperature differentials similar to those in full-scale systems, and can simulate full-scale wall losses. Surface area to volume ratios, a significant factor in terms of heat loss through vessel walls, were estimated by the present authors at 5.0-88.0 m 2 /m 3 for experimental composting reactors and 0.4-3.8 m 2 /m 3 for full-scale systems. Non-thermodynamic factors such as compression, sidewall airflow effects, channelling and mixing may affect simulation performance and are discussed. Further work to investigate wall effects in composting reactors, to obtain more data on horizontal temperature profiles and rates of biological heat production, to incorporate compressive effects into experimental reactors and to investigate experimental systems employing natural ventilation is suggested

  12. Responses of the biogas process to pulses of oleate in reactors treating mixtures of cattle and pig manure.

    Science.gov (United States)

    Nielsen, Henrik Bangsø; Ahring, Birgitte Kiaer

    2006-09-05

    The effect of oleate on the anaerobic digestion process was investigated. Two thermophilic continuously stirred tank reactors (CSTR) were fed with mixtures of cattle and pig manure with different total solid (TS) and volatile solid (VS) content. The reactors were subjected to increasing pulses of oleate. Following pulses of 0.5 and 1.0 g oleate/L, the most distinct increase in volatile fatty acid (VFA) concentrations were observed in the reactor with the lowest TS/VS content. This suggests a higher adsorption of oleate on the surfaces of biofibers in the reactor with the highest TS/VS and a less pronounced inhibition of the anaerobic digestion process. On the other hand, addition of 2.0 g oleate/L severely inhibited the process in both reactors, and a significant increase in all VFA concentrations combined with an immediate drop in methane production was noticed. However, 20 days after the reactors had been exposed to oleate both reactors showed a lower VFA concentration along with a higher methane production than before the pulses. This indicates that oleate had a stimulating effect on the overall process. The improved acetogenic and methanogenic activity in the reactors was confirmed in batch activity tests. In addition to this, toxicity tests revealed that the oleate pulses induced an increase in the tolerance level of acetotrophic methanogens towards oleate. When evaluating the usability of different process parameters (i.e., VFA and methane production) as indicators of process recovery, following the inhibition by oleate, propionate was found to be most suitable. (c) 2006 Wiley Periodicals, Inc.

  13. Removal of Total Coliforms, Thermotolerant Coliforms, and Helminth Eggs in Swine Production Wastewater Treated in Anaerobic and Aerobic Reactors

    OpenAIRE

    Silvia Helena Zacarias Sylvestre; Estevam Guilherme Lux Hoppe; Roberto Alves de Oliveira

    2014-01-01

    The present work evaluated the performance of two treatment systems in reducing indicators of biological contamination in swine production wastewater. System I consisted of two upflow anaerobic sludge blanket (UASB) reactors, with 510 and 209 L in volume, being serially arranged. System II consisted of a UASB reactor, anaerobic filter, trickling filter, and decanter, being also organized in series, with volumes of 300, 190, 250, and 150 L, respectively. Hydraulic retention times (HRT) applied...

  14. Variational techniques for reactor physics calculations of heterogeneous reactor cores. Final report for April 15, 1993--April 14, 1995

    International Nuclear Information System (INIS)

    Wojtowicz, G.M.; Holloway, J.P.

    1995-06-01

    Variational coarse mesh techniques are developed for the solution of the one group neutron transport equation in one-dimensional reactor lattices. In contrast to conventional nodal lattice applications which discretize diffusion theory and use node homogenized cross sections, the authors retain the spatial dependence of the cross sections and instead employ an alternative flux representation. The initial form of this flux representation (trial function) for the angular flux was inspired by the leading order solution in the asymptotic expansion of the angular flux--namely, the slow modulation of a periodic pin cell flux. The authors called the variational technique based on this form of trial function the Spectral Element Asymptotic Method (SEAM); it is capable of achieving order of magnitude reductions of eigenvalue and pointwise scalar flux errors as compared with diffusion theory. A different trial function can be developed based on the leading order and first order correction terms in the asymptotic expansion of the angular flux. SuperSEAM, the method based on this new trial function, allows the neutron transport equation to be cast into a form whose solution has much slower spatial variation than the SEAM solution; thus, the SuperSEAM result can be accurately described with fewer variables. SuperSEAM is therefore capable of achieving the same high degree of accuracy as SEAM at a cost comparable with homogenized nodal diffusion theory

  15. Physical-chemical effects of irrigation with treated wastewater on Dusky Red Latosol soil

    Directory of Open Access Journals (Sweden)

    Vanessa Ribeiro Urbano

    2015-11-01

    Full Text Available The current water crisis underlines the importance of improving water management. The use of effluent from secondary treatment in agriculture can reduce the discharge of effluent into natural bodies and provide nutrients to crops. This study evaluated the physical and chemical properties of a Dusky Red Latosol soil that had been irrigated with treated wastewater. Conducted at the Center of Agricultural Sciences (CCA of Federal University of São Carlos (UFSCar, in Araras/São Paulo/Brazil, 18 undisturbed soil samples were collected and deposited on a constant-head permeameter in order to simulate the irrigation of five growth cycles of lettuce (Lactuca sativa L., organized in five different treatments and one control group. For each treatment 0.58 L, 1.16 L, 1.74 L, 2.32 L, and 2.90 L of treated wastewater and distilled water were applied . The treated wastewater came from a domestic waste treatment plant. After the water filtered through the soil, samples of treated wastewater were collected for analyses of electrical conductivity (EC, sodium adsorption ratio (SAR, turbidity, pH, Na, K, Mg, P and Ca and, in the soil the granulometry, complete fertility, exchangeable sodium percentage (ESP and saturated hydraulic conductivity (Ksat. The Ksat decreased, but did not alter the infiltration of water and nutrients in the soil. The concentration of nutrients in the soil increased, including Na, which raises the need for monitoring soil’s salinity. In conclusion, the application of wastewater did not cause damage to the physical properties of the soil, but resulted in a tendency towards salinization.

  16. Characterization samples of Tigris river water treated with nano colloidal silver (physically, chemically, microbiologically)

    International Nuclear Information System (INIS)

    Dumboos, H. I.; Beden, S. J.; Zouari, K.; Chkir, N.; Ahmed, H. A.

    2012-12-01

    Many researches of using nano silver in purification of drinking water from bacteria and its effect on stan dared properties as drinking water were established. Two stages accomplished in these projects. First stage include preparation of colloidal silver with characterization process and prepare water samples through sedimentation, filtration process, PH and turbidity measure then treated with colloidal silver in volume ratio (0.1-Λ) ml/100ml. The second stage represent select the better results from stage one and take samples to determine the standard characterization values with chemical, physical and microbiological taste. Results will be compared with Iraq standard certification. (Author)

  17. Performance evaluation of an anaerobic fluidized bed reactor with natural zeolite as support material when treating high-strength distillery wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, N. [Renewable Energy Technology Center (CETER), ' ' Jose Antonio Echeverria' ' Polytechnical University, Calle 127 s/n, CP 19390, Apdo. 6028, Habana 6 Marianao, Ciudad de La Habana (Cuba); Montalvo, S. [Department of Chemical Engineering, Santiago de Chile University, Ave. Lib. Bernardo O' Higgins 3363, Santiago de Chile (Chile); Borja, R.; Travieso, L.; Raposo, F. [Instituto de la Grasa (CSIC), Avenida Padre Garcia Tejero 4, 41012 Sevilla (Spain); Guerrero, L. [Department of Chemical, Biotechnological and Environmental Processes, Federico Santa Maria Technical University, Casilla 110-V, Valparaiso (Chile); Sanchez, E.; Colmenarejo, M.F. [Centro de Ciencias Medioambientales (CSIC), C/Serrano, 115-Duplicado, 28006 Madrid (Spain); Cortes, I. [Environment Nacional Center, Chile University, Ave. Larrain 9975, La Reina, Santiago de Chile (Chile)

    2008-11-15

    The performance of two laboratory-scale fluidized bed reactors with natural zeolite as support material when treating high-strength distillery wastewater was assessed. Two sets of experiments were carried out. In the first experimental set, the influences of the organic loading rate (OLR), the fluidization level (FL) and the particle diameter of the natural zeolite (D{sub P}) were evaluated. This experimental set was carried out at an OLR from 2 to 5 g COD (chemical oxygen demand)/l d, at FL 20% and 40% and with D{sub P} in the range of 0.2-0.5 mm (reactor 1) and of 0.5-0.8 mm (reactor 2). It was demonstrated that OLR and FL had a slight influence on COD removal, whereas they had a strong influence on the methane production rate. The COD removal was slightly higher for the highest particle diameter used. The second experimental set was carried out at an OLR from 3 to 20 g COD/l d with 25% of fluidization and D{sub P} in the above-mentioned ranges for reactors 1 and 2. The performance of the two reactors was similar; no significant differences were found. The COD removal efficiency correlated with the OLR based on a straight line. COD removal efficiencies higher than 80% were achieved in both reactors without significant differences. In addition, a straight line equation with a slope of 1.74 d{sup -1} and an intercept on the y-axis equal to zero described satisfactorily the effect of the influent COD on the COD removal rate. It was also observed that both COD removal rate and methane production (Q{sub M}) increased linearly with the OLR, independently of the D{sub P} used. (author)

  18. Influence of biomass acclimation on the performance of a partial nitritation-anammox reactor treating industrial saline effluents.

    Science.gov (United States)

    Giustinianovich, Elisa A; Campos, José-Luis; Roeckel, Marlene D; Estrada, Alejandro J; Mosquera-Corral, Anuska; Val Del Río, Ángeles

    2018-03-01

    The performance of the partial nitritation/anammox processes was evaluated for the treatment of fish canning effluents. A sequencing batch reactor (SBR) was fed with industrial wastewater, with variable salt and total ammonium nitrogen (TAN) concentrations in the range of 1.75-18.00 g-NaCl L -1 and 112 - 267 mg-TAN L -1 . The SBR operation was divided into two experiments: (A) progressive increase of salt concentrations from 1.75 to 18.33 g-NaCl L -1 ; (B) direct application of high salt concentration (18 g-NaCl L -1 ). The progressive increase of NaCl concentration provoked the inhibition of the anammox biomass by up to 94% when 18 g-NaCl L -1 were added. The stable operation of the processes was achieved after 154 days when the nitrogen removal rate was 0.021 ± 0.007 g N/L·d (corresponding to 30% of removal efficiency). To avoid the development of NOB activity at low salt concentrations and to stabilize the performance of the processes dissolved oxygen was supplied by intermittent aeration. A greater removal rate of 0.029 ± 0.017 g-N L -1 d -1 was obtained with direct exposure of the inoculum to 18 g-NaCl L -1 in less than 40 days. Also, higher specific activities than those from the inoculum were achieved for salt concentrations of 15 and 20 g-NaCl L -1 after 39 days of operation. This first study of the performance of the partial nitritation/anammox processes, to treat saline wastewaters, indicates that the acclimation period can be avoided to shorten the start-up period for industrial application purposes. Nevertheless, further experiments are needed in order to improve the efficiency of the processes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    International Nuclear Information System (INIS)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W.

    2016-01-01

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  20. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-05-15

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  1. Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU. Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.

  2. FN approximation of the solution to a singular integral equation of classical reactor physics

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    2004-01-01

    The iterated FN method is applied to a singular integral equation arising from a classical problem of reactor physics to determine the distribution of fissile material giving a spatially uniform flux. The FN iterations are accelerated toward convergence through the Wynn-algorithm - but first - Happy Birthday 'Fast Eddie' Larsen Why do I refer to the well known, most proper and exquisitely accomplished Edward W. Larsen as 'Fast Eddie'. Well our story begins in a small back bar room in the lobby of one of Los Alamos' finest and most luxurious hotels. Two young men were having a transport theoretic discussion while they were engaged in a serious game of pool with monetary benefits going to the winner. In addition, the two were sipping their most favorite lavation in rather large quantities - one, a short stocky man with thinning hair, was sipping to forget the cost of his recent divorce, and the other, a shorter stockier man also with thinning hair, was drinking, well because he liked to drink and it just made him silly. As they continued their transport discussion, one stocky man turned to the other and said, 'I wonder what 'Fast Eddie' Larsen would say to our transport question'. The other stocky man immediately thought the 'Fast Eddie' reference was to Paul Newman who played 'Fast Eddie', an expert at applied particle transport theory (a pool player) in the movie the Hustler and asked if indeed this was the case. The first stocky man said 'No. I call everyone with the name Ed 'Fast Eddie' ' - and that's the story of how 'Fast Eddie' Larsen got his name. Happy 60th Ed and thanks for all the great transport theory - from one of your biggest fans

  3. FN approximation of the solution to a singular integral equation of classical reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Ganapol, B.D. [Department of Aerospace and Mechanical Engineering, University of Arizona, AME Building, Tucson, AZ 85721 (United States)]. E-mail: ganapol@ame.arizona.edu

    2004-11-01

    The iterated FN method is applied to a singular integral equation arising from a classical problem of reactor physics to determine the distribution of fissile material giving a spatially uniform flux. The FN iterations are accelerated toward convergence through the Wynn-algorithm - but first - Happy Birthday 'Fast Eddie' Larsen Why do I refer to the well known, most proper and exquisitely accomplished Edward W. Larsen as 'Fast Eddie'. Well our story begins in a small back bar room in the lobby of one of Los Alamos' finest and most luxurious hotels. Two young men were having a transport theoretic discussion while they were engaged in a serious game of pool with monetary benefits going to the winner. In addition, the two were sipping their most favorite lavation in rather large quantities - one, a short stocky man with thinning hair, was sipping to forget the cost of his recent divorce, and the other, a shorter stockier man also with thinning hair, was drinking, well because he liked to drink and it just made him silly. As they continued their transport discussion, one stocky man turned to the other and said, 'I wonder what 'Fast Eddie' Larsen would say to our transport question'. The other stocky man immediately thought the 'Fast Eddie' reference was to Paul Newman who played 'Fast Eddie', an expert at applied particle transport theory (a pool player) in the movie the Hustler and asked if indeed this was the case. The first stocky man said 'No. I call everyone with the name Ed 'Fast Eddie' ' - and that's the story of how 'Fast Eddie' Larsen got his name. Happy 60th Ed and thanks for all the great transport theory - from one of your biggest fans.

  4. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  5. Progress and challenges in the development and qualification of multi-level multi-physics coupled methodologies for reactor analysis

    International Nuclear Information System (INIS)

    Ivanov, K.; Avramova, M.

    2007-01-01

    Current trends in nuclear power generation and regulation as well as the design of next generation reactor concepts along with the continuing computer technology progress stimulate the development, qualification and application of multi-physics multi-scale coupled code systems. The efforts have been focused on extending the analysis capabilities by coupling models, which simulate different phenomena or system components, as well as on refining the scale and level of detail of the coupling. This paper reviews the progress made in this area and outlines the remaining challenges. The discussion is illustrated with examples based on neutronics/thermohydraulics coupling in the reactor core modeling. In both fields recent advances and developments are towards more physics-based high-fidelity simulations, which require implementation of improved and flexible coupling methodologies. First, the progresses in coupling of different physics codes along with the advances in multi-level techniques for coupled code simulations are discussed. Second, the issues related to the consistent qualification of coupled multi-physics and multi-scale code systems for design and safety evaluation are presented. The increased importance of uncertainty and sensitivity analysis are discussed along with approaches to propagate the uncertainty quantification between the codes. The incoming OECD LWR Uncertainty Analysis in Modeling (UAM) benchmark is the first international activity to address this issue and it is described in the paper. Finally, the remaining challenges with multi-physics coupling are outlined. (authors)

  6. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

  7. The Text of the Agreement between the Agency and the Governments of Norway, Poland and Yugoslavia concerning Co-operative Research in Reactor Physics

    International Nuclear Information System (INIS)

    1964-01-01

    The text of the Agreement between the Agency and the Governments of Norway, Poland and Yugoslavia concerning Co-operative Research in Reactor Physics (the 'NPY Agreement') is reproduced in this document for the information of all Members

  8. Physical-chemistry problems in safe disposal of irradiated fuel from RA research reactor

    International Nuclear Information System (INIS)

    Spent fuel resulting from 25 years of operating the 6.5/10 MW heavy water moderated and cooled research RA at the Vinca Institute is still all stored in the temporary spent fuel storage pool in the basement of the reactor building. In 1984, the reactor was shut down for refurbishment, which for a number of reasons has not yet been completed. However, independently of the future status of the research reactor, safe disposal of the so far irradiated fuel must be the subject of primary concern. Basic facts about operation, ageing, reconstruction and spent fuel storage of the research reactor RA have been presented and discussed in detail in some earlier papers. This paper describes present activities and discusses options for permanent solution of the spent fuel storage problem. (author)

  9. 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)

  10. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    OpenAIRE

    Winter, Dominik

    2014-01-01

    The economical operation of a boiling water reactor (BWR) is mainly achieved by the axially uniform utilization of the nuclear fuel in the assemblies which is challenging because the neutron spectrum in the active reactor core varies with the axial position. More precisely, the neutron spectrum becomes harder the higher the position is resulting in a decrease of the fuel utilization because the microscopic fission cross section is smaller by several orders of magnitude. In this work, the use ...

  11. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  12. Development and evaluation of a radial anaerobic/aerobic reactor treating organic matter and nitrogen in sewage

    Directory of Open Access Journals (Sweden)

    L. H. P. Garbossa

    2005-12-01

    Full Text Available The design and performance of a radial anaerobic/aerobic immobilized biomass (RAAIB reactor operating to remove organic matter, solids and nitrogen from sewage are discussed. The bench-scale RAAIB was divided into five concentric chambers. The second and fourth chambers were packed with polyurethane foam matrices. The performance of the reactor in removing organic matter and producing nitrified effluent was good, and its configuration favored the transfer of oxygen to the liquid mass due to its characteristics and the fixed polyurethane foam bed arrangement in concentric chambers. Partial denitrification of the liquid also took place in the RAAIB. The reactor achieved an organic matter removal efficiency of 84%, expressed as chemical oxygen demand (COD, and a total Kjeldahl nitrogen (TKN removal efficiency of 96%. Average COD, nitrite and nitrate values for the final effluent were 54 mg.L-1, 0.3 mg.L-1 and 22.1 mg.L-1, respectively.

  13. Physics and safety studies of a low conversion ratio sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Smith, M. A.; Hill, R. N.; Dunn, F. E.

    2004-01-01

    This paper explores the feasibility of a compact fast burner reactor that can achieve a low transuranic conversion ratio. The major design option considered is the reduction of fissile breeding by the removal of fertile material from the fast reactor system. Reductions in the fuel pin diameter and thus fuel loading were employed to remove fertile material. Reactor performance parameters and reactivity coefficients were evaluated for a compact core design with a targeted conversion ratio of 0.25. To assess the safety implications, a detailed transient analysis model was employed using the SAS4A/SASSYS-1 computer code. A series of calculations was performed to assess the behavior of the reactor and plant in an unprotected loss-of-flow accident (ULOF). A parametric study was also carried out using increasingly conservative modeling assumptions. The computational results show that for nominal, best-estimate analysis assumptions and input data, the low conversion ratio reactor design responds to the ULOF with a very high level of self-protection. Both short-term and long-term quasi-equilibrium reactor conditions predicted in the analysis indicate very large margins of safety. (authors)

  14. A fast linear predictive adaptive model of packed bed coupled with UASB reactor treating onion waste to produce biofuel.

    Science.gov (United States)

    Milquez-Sanabria, Harvey; Blanco-Cocom, Luis; Alzate-Gaviria, Liliana

    2016-10-03

    Agro-industrial wastes are an energy source for different industries. However, its application has not reached small industries. Previous and current research activities performed on the acidogenic phase of two-phase anaerobic digestion processes deal particularly with process optimization of the acid-phase reactors operating with a wide variety of substrates, both soluble and complex in nature. Mathematical models for anaerobic digestion have been developed to understand and improve the efficient operation of the process. At present, lineal models with the advantages of requiring less data, predicting future behavior and updating when a new set of data becomes available have been developed. The aim of this research was to contribute to the reduction of organic solid waste, generate biogas and develop a simple but accurate mathematical model to predict the behavior of the UASB reactor. The system was maintained separate for 14 days during which hydrolytic and acetogenic bacteria broke down onion waste, produced and accumulated volatile fatty acids. On this day, two reactors were coupled and the system continued for 16 days more. The biogas and methane yields and volatile solid reduction were 0.6 ± 0.05 m 3 (kg VS removed ) -1 , 0.43 ± 0.06 m 3 (kg VS removed ) -1 and 83.5 ± 9.8 %, respectively. The model application showed a good prediction of all process parameters defined; maximum error between experimental and predicted value was 1.84 % for alkalinity profile. A linear predictive adaptive model for anaerobic digestion of onion waste in a two-stage process was determined under batch-fed condition. Organic load rate (OLR) was maintained constant for the entire operation, modifying effluent hydrolysis reactor feed to UASB reactor. This condition avoids intoxication of UASB reactor and also limits external buffer addition.

  15. Multi-physic simulations of irradiation experiments in a technological irradiation reactor; Modelisation pluridisciplinaire d'experiences d'irradiation dans un reacteur d'irradiation technologique

    Energy Technology Data Exchange (ETDEWEB)

    Bonaccorsi, Th

    2007-09-15

    A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

  16. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    Energy Technology Data Exchange (ETDEWEB)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

  17. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    International Nuclear Information System (INIS)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-01-01

    Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR-06 are highlighted, and the future of the two projects is discussed

  18. Supervised physical therapy in women treated with radiotherapy for breast cancer

    Directory of Open Access Journals (Sweden)

    Nara Fernanda Braz da Silva Leal

    Full Text Available ABSTRACT Objective: to evaluate the effect of physical therapy on the range of motion of the shoulders and perimetry of the upper limbs in women treated with radiotherapy for breast cancer. Methods: a total of 35 participants were randomized into two groups, with 18 in the control group (CG and 17 in the study group (SG. Both of the groups underwent three evaluations to assess the range of motion of the shoulders and perimetry of the upper limbs, and the study group underwent supervised physical therapy for the upper limbs. Results: the CG had deficits in external rotation in evaluations 1, 2, and 3, whereas the SG had deficits in flexion, abduction, and external rotation in evaluation 1. The deficit in abduction was recovered in evaluation 2, whereas the deficits in all movements were recovered in evaluation 3. No significant differences in perimetry were observed between the groups. Conclusion: the applied supervised physical therapy was effective in recovering the deficit in abduction after radiotherapy, and the deficits in flexion and external rotation were recovered within two months after the end of radiotherapy. Registration number of the clinical trial: NCT02198118.

  19. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  20. Geometrical modification transfer between specific meshes of each coupled physical codes. Application to the Jules Horowitz research reactor experimental devices

    International Nuclear Information System (INIS)

    Duplex, B.

    2011-01-01

    The CEA develops and uses scientific software, called physical codes, in various physical disciplines to optimize installation and experimentation costs. During a study, several physical phenomena interact, so a code coupling and some data exchanges between different physical codes are required. Each physical code computes on a particular geometry, usually represented by a mesh composed of thousands to millions of elements. This PhD Thesis focuses on the geometrical modification transfer between specific meshes of each coupled physical code. First, it presents a physical code coupling method where deformations are computed by one of these codes. Next, it discusses the establishment of a model, common to different physical codes, grouping all the shared data. Finally, it covers the deformation transfers between meshes of the same geometry or adjacent geometries. Geometrical modifications are discrete data because they are based on a mesh. In order to permit every code to access deformations and to transfer them, a continuous representation is computed. Two functions are developed, one with a global support, and the other with a local support. Both functions combine a simplification method and a radial basis function network. A whole use case is dedicated to the Jules Horowitz reactor. The effect of differential dilatations on experimental device cooling is studied. (author) [fr

  1. Optimization of the physical characteristics and the regime of uniform partial refuelings of a fast nuclear reactor

    Science.gov (United States)

    Kuz'min, A. M.; Moroko, V. I.

    2013-12-01

    This paper covers some specific features of the optimization problem with integer-valued and continuously changing parameters that has been formulated for a fast reactor operating under the steady-state regime of the uniform partial refueling. Effective algorithms for calculating the physical characteristics and an iterative procedure of constructing optimum values of parameters are proposed. The paper considers the solution of a problem on minimization of the loss of energy generation in a reactor of the BREST-800 type that occurs because average fuel burnup in fuel assemblies being removed does not achieve its maximum permissible level. For several core arrangements, the comparison with nonoptimum solutions is given and the role of various factors contributing to an increase in average fuel burnup is evaluated.

  2. Reactor based plutonium disposition - physics and fuel behaviour benchmark studies of an OECD/NEA experts group

    International Nuclear Information System (INIS)

    D'Hondt, P.; Gehin, J.; Na, B.C.; Sartori, E.; Wiesenack, W.

    2001-01-01

    One of the options envisaged for disposing of weapons grade plutonium, declared surplus for national defence in the Russian Federation and Usa, is to burn it in nuclear power reactors. The scientific/technical know-how accumulated in the use of MOX as a fuel for electricity generation is of great relevance for the plutonium disposition programmes. An Expert Group of the OECD/Nea is carrying out a series of benchmarks with the aim of facilitating the use of this know-how for meeting this objective. This paper describes the background that led to establishing the Expert Group, and the present status of results from these benchmarks. The benchmark studies cover a theoretical reactor physics benchmark on a VVER-1000 core loaded with MOX, two experimental benchmarks on MOX lattices and a benchmark concerned with MOX fuel behaviour for both solid and hollow pellets. First conclusions are outlined as well as future work. (author)

  3. Effect of hydraulic retention time on hydrodynamic behavior of anaerobic-aerobic fixed bed reactor treating cattle slaughterhouse effluent

    Directory of Open Access Journals (Sweden)

    Daiane Cristina de Freitas

    2017-09-01

    Full Text Available The study of the hydrodynamic behavior in reactors provides characteristics of the flow regime and its anomalies that can reduce biological processes efficiency due to the decrease of the useful volume and the hydraulic retention time required for the performance of microbial activity. In this study, the hydrodynamic behavior of an anaerobic-aerobic fixed bed reactor, operated with HRT (hydraulic retention time of 24, 18 and 12 hours, was evaluated in the treatment of raw cattle slaughterhouse wastewater. Polyurethane foam and expanded clay were used as support media for biomass immobilization. Experimental data of pulse type stimulus-response assays were performed with eosin Y and bromophenol blue, and adjusted to the single-parameter theoretical models of dispersion and N-continuous stirred tank reactors in series (N-CSTR. N-CSTR model presented the best adjustment for the HRT and tracers evaluated. RDT (residence time distribution curves obtained with N-CSTR model in the assays with bromophenol blue resulted in better adjustment compared to the eosin Y. The predominant flow regime in AAFBR (anaerobic aerobic fixed bed reactor is the N-CSTR in series, as well as the existence of preferential paths and hydraulic short-circuiting.

  4. Functionally redundant but dissimilar microbial communities within biogas reactors treating maize silage in co-fermentation with sugar beet silage

    Science.gov (United States)

    Langer, Susanne G; Ahmed, Sharif; Einfalt, Daniel; Bengelsdorf, Frank R; Kazda, Marian

    2015-01-01

    Numerous observations indicate a high flexibility of microbial communities in different biogas reactors during anaerobic digestion. Here, we describe the functional redundancy and structural changes of involved microbial communities in four lab-scale continuously stirred tank reactors (CSTRs, 39°C, 12 L volume) supplied with different mixtures of maize silage (MS) and sugar beet silage (SBS) over 80 days. Continuously stirred tank reactors were fed with mixtures of MS and SBS in volatile solid ratios of 1:0 (Continuous Fermenter (CF) 1), 6:1 (CF2), 3:1 (CF3), 1:3 (CF4) with equal organic loading rates (OLR 1.25 kgVS m−3 d−1) and showed similar biogas production rates in all reactors. The compositions of bacterial and archaeal communities were analysed by 454 amplicon sequencing approach based on 16S rRNA genes. Both bacterial and archaeal communities shifted with increasing amounts of SBS. Especially pronounced were changes in the archaeal composition towards Methanosarcina with increasing proportion of SBS, while Methanosaeta declined simultaneously. Compositional shifts within the microbial communities did not influence the respective biogas production rates indicating that these communities adapted to environmental conditions induced by different feedstock mixtures. The diverse microbial communities optimized their metabolism in a way that ensured efficient biogas production. PMID:26200922

  5. Removal of Total Coliforms, Thermotolerant Coliforms, and Helminth Eggs in Swine Production Wastewater Treated in Anaerobic and Aerobic Reactors

    Science.gov (United States)

    Zacarias Sylvestre, Silvia Helena; Lux Hoppe, Estevam Guilherme; de Oliveira, Roberto Alves

    2014-01-01

    The present work evaluated the performance of two treatment systems in reducing indicators of biological contamination in swine production wastewater. System I consisted of two upflow anaerobic sludge blanket (UASB) reactors, with 510 and 209 L in volume, being serially arranged. System II consisted of a UASB reactor, anaerobic filter, trickling filter, and decanter, being also organized in series, with volumes of 300, 190, 250, and 150 L, respectively. Hydraulic retention times (HRT) applied in the first UASB reactors were 40, 30, 20, and 11 h in systems I and II. The average removal efficiencies of total and thermotolerant coliforms in system I were 92.92% to 99.50% and 94.29% to 99.56%, respectively, and increased in system II to 99.45% to 99.91% and 99.52% to 99.93%, respectively. Average removal rates of helminth eggs in system I were 96.44% to 99.11%, reaching 100% as in system II. In reactor sludge, the counts of total and thermotolerant coliforms ranged between 105 and 109 MPN (100 mL)−1, while helminth eggs ranged from 0.86 to 9.27 eggs g−1 TS. PMID:24812560

  6. Effect Of Organic Substrate Composition On Microbial Community Structure Of Pilot-Scale Biochemical Reactors Treating Mining Influenced Water

    Science.gov (United States)

    Mining-influenced water (MIW) is acidic, metal rich water formed when sulfide minerals react with oxygen and water. There are various options for the treatment of MIW; however, passive biological systems such as biochemical reactors (BCRs) have shown promise because of their low...

  7. Effect Of Organic Substrate Composition On Microbial Community Structure Of Pilot-Scale Biochemical Reactors Treating Mining Influenced Water - (Presentation)

    Science.gov (United States)

    Mining-influenced water (MIW) is acidic, metal rich water formed when sulfide minerals react with oxygen and water. There are various options for the treatment of MIW; however, passive biological systems such as biochemical reactors (BCRs) have shown promise because of their low...

  8. Removal of total coliforms, thermotolerant coliforms, and helminth eggs in Swine production wastewater treated in anaerobic and aerobic reactors.

    Science.gov (United States)

    Zacarias Sylvestre, Silvia Helena; Lux Hoppe, Estevam Guilherme; de Oliveira, Roberto Alves

    2014-01-01

    The present work evaluated the performance of two treatment systems in reducing indicators of biological contamination in swine production wastewater. System I consisted of two upflow anaerobic sludge blanket (UASB) reactors, with 510 and 209 L in volume, being serially arranged. System II consisted of a UASB reactor, anaerobic filter, trickling filter, and decanter, being also organized in series, with volumes of 300, 190, 250, and 150 L, respectively. Hydraulic retention times (HRT) applied in the first UASB reactors were 40, 30, 20, and 11 h in systems I and II. The average removal efficiencies of total and thermotolerant coliforms in system I were 92.92% to 99.50% and 94.29% to 99.56%, respectively, and increased in system II to 99.45% to 99.91% and 99.52% to 99.93%, respectively. Average removal rates of helminth eggs in system I were 96.44% to 99.11%, reaching 100% as in system II. In reactor sludge, the counts of total and thermotolerant coliforms ranged between 10(5) and 10(9) MPN (100 mL)(-1), while helminth eggs ranged from 0.86 to 9.27 eggs g(-1) TS.

  9. Aspects of the physics and chemistry of water radiolysis by fast neutrons and fast electrons in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCracken, D.R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Tsang, K.T. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Laughton, P.J

    1998-09-01

    Detailed radiation physics calculations of energy deposition have been done for the coolant of CANDU reactors and Pressurized Water Reactors (PWRs). The geometry of the CANDU fuel channel was modelled in detail. Fluxes and energy-deposition rates for neutrons, recoil ions, photons, and fast electrons have been calculated using MCNP4B, WIMS-AECL, and specifically derived energy-transfer factors. These factors generate the energy/flux spectra of recoil ions from fast-neutron energy/flux spectra. The energy spectrum was divided into 89 discrete ranges (energy bins).The production of oxidizing species and net coolant radiolysis can be suppressed by the addition of hydrogen to the coolant of nuclear reactors. It is argued that the net dissociation of coolant by gamma rays is suppressed by lower levels of excess hydrogen than when dissociation is by ion recoils. This has consequences for the modelling of coolant radiolysis by homogeneous kinetics. More added hydrogen is required to stop water radiolysis by recoil ions acting alone than if recoil ions and gamma rays acted concurrently in space and time. Homogeneous kinetic models and experimental data suggest that track overlap is very inefficient in providing radicals from gamma-ray tracks to recombine molecular products in ion-recoil tracks. An inhomogeneous chemical model is needed that incorporates ionizing-particle track structure and track overlap. Such a model does not yet exist, but a number of limiting cases using homogeneous kinetics are discussed. There are sufficient uncertainties and contradictions in the data relevant to the radiolysis of reactor coolant that the relatively high CHC's (critical hydrogen concentration) observed in NRU reactor experiments (compared to model predictions) may be explainable by errors in fundamental data and understanding of water radiolysis under reactor conditions. The radiation chemistry program at CRL has been focused to generate quantitative water-radiolysis data in a

  10. Advanced core physics and thermal hydraulics analysis of boiling water reactors using innovative fuel concepts

    International Nuclear Information System (INIS)

    Winter, Dominik

    2014-01-01

    increase of the fission reaction rate of plutonium in the upper part of the active height leading to an increase of the neutron flux and subsequent induced fission reactions - ThPu: an increase of the fissile inventory ratio with ongoing operation due to the production of the fissile uranium-233 from the fertile thorium-232 Finally, for both core loadings, equal or even more favorable reactor physical safety parameters are achieved which remain within the licensed safety margins of today's BWRs.

  11. Physics studies of weapons plutonium disposition in the Integral Fast Reactor closed fuel cycle

    International Nuclear Information System (INIS)

    Hill, R.N.; Wade, D.C.; Liaw, J.R.; Fujita, E.K.

    1995-01-01

    The core performance impact of weapons plutonium introduction into the Integral Fast Reactor (IFR) closed fuel cycle is investigated by comparing three disposition scenarios: a power production mode, a moderate destruction mode, and a maximum destruction mode, all at a constant heat rating of 840 MW(thermal). For each scenario, two fuel cycle models are evaluated: cores using weapons material as the sole source of transuranics in a once-through mode and recycle cores using weapons material only as required for a makeup feed. In addition, the impact of alternative feeds (recycled light water reactor or liquid-metal reactor transuranics) on burner core performance is assessed. Calculated results include mass flows, detailed isotopic distributions, neutronic performance characteristics, and reactivity feedback coefficients. In general, it is shown that weapons plutonium does not have an adverse effect on IFR core performance characteristics; also, favorable performance can be maintained for a wide variety of feed materials and fuel cycle strategies

  12. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  13. Physical education class injuries treated in emergency departments in the US in 1997-2007.

    Science.gov (United States)

    Nelson, Nicolas G; Alhajj, Maya; Yard, Ellen; Comstock, Dawn; McKenzie, Lara B

    2009-09-01

    The goal was to describe the epidemiological features of physical education (PE)-related injuries treated in US emergency departments. A retrospective analysis was conducted with data for children and adolescents (5-18 years of age) from the National Electronic Injury Surveillance Study of the US Consumer Product Safety Commission, from 1997 through 2007. Sample weights provided by the National Electronic Injury Surveillance System were used to calculate national estimates of PE-related injuries. Trend significance of the number of PE-related injuries over time was analyzed by using linear regression analysis. An estimated 405305 children and adolescents were treated in emergency departments for PE-related injuries. The annual number of cases increased 150% during the study period (P = .001). Nearly 70% of PE-related injuries occurred during 6 activities, that is, running, basketball, football, volleyball, soccer, and gymnastics. Boys' injuries were more likely to involve the head, to be diagnosed as a laceration or fracture, to be attributable to contact with a person or structure, and to occur during group activities. Girls' injuries were more likely to involve the lower extremities, to be strains and sprains, to be acute noncontact injuries, and to occur during individual activities. More research is needed to identify the cause of the increase in PE-related injuries, to examine the gender difference in PE-related injuries, and to determine appropriate injury prevention solutions and policies.

  14. Effect of plasma physics on choices of first wall materials and structures for a thermonuclear reactor

    International Nuclear Information System (INIS)

    Meade, D.M.

    1975-01-01

    Impurity ions adversely affect the behavior of present-day tokamaks, and control of impurities is expected to be a key element in determining the feasibility of thermonuclear fusion reactors. The plasma-surface interactions for tokamaks and several techniques for controlling impurities are described. The plasma-surface problem of next generation devices PLT, PDX, DIII and TFTR is expected to be similar to those encountered in a reactor. For these devices calculations indicate that most of the particle energy efflux will be in the 1 keV region. Ironically this energy region has not yet been investigated thoroughly by the surface physicists

  15. Use of Physical Therapists to Identify and Treat Musculoskeletal Injuries at "The Tip of the Trident".

    Science.gov (United States)

    Shaw, Jesse; Brown, Laura; Jansen, Brittany

    Musculoskeletal injuries continue to be the most common cause of decreased readiness and loss of productivity in all military environments. In commands with smaller footprints, such as Naval Special Warfare (NSW), every asset is critical for mission success. Studies have shown that early intervention by a medical provider can enhance healing and maintain unit readiness by preventing medical evacuations. Reports are limited with regard to Special Forces commands, especially during deployment. This article describes the injury characteristics and treatment of injuries seen by a physical therapist while deployed at forward operation commands embedded with NSW Group 2 Team 4. Over 4 months, 282 patients were evaluated and treated in southeast Afghanistan. In descending order, the three most common injured body regions were the lumbar/sacral spine (n = 82), shoulder (n = 59), and knee (n = 28). Therapy exercises (n = 461) were the most frequently performed treatment modality, followed by mobilization/manipulation (n = 394) and dry needling (n = 176). No patient evaluated was medically evacuated from the area or sent to an advanced medical site. Our data are similar to other published data reported on deployed units in terms of mechanisms and locations of injuries; thus, Special Forces commands do not appear to have unique injury patterns. These results support continued use of physical therapists in forward operations because of their ability to evaluate injuries and provide treatment modalities that help maintain the integrity of small commands at the site of injury. 2017.

  16. Physical Analysis of the Initial Core and Running-In Phase for Pebble-Bed Reactor HTR-PM

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2017-01-01

    Full Text Available The pebble-bed reactor HTR-PM is being built in China and is planned to be critical in one or two years. At present, one emphasis of engineering design is to determine the fuel management scheme of the initial core and running-in phase. There are many possible schemes, and many factors need to be considered in the process of scheme evaluation and analysis. Based on the experience from the constructed or designed pebble-bed reactors, the fuel enrichment and the ratio of fuel spheres to graphite spheres are important. In this paper, some relevant physical considerations of the initial core and running-in phase of HTR-PM are given. Then a typical scheme of the initial core and running-in phase is proposed and simulated with VSOP code, and some key physical parameters, such as the maximum power per fuel sphere, the maximum fuel temperature, the refueling rate, and the discharge burnup, are calculated. Results of the physical parameters all satisfy the relevant design requirements, which means the proposed scheme is safe and reliable and can provide support for the fuel management of HTR-PM in the future.

  17. Expert systems for the analysis of transients on nuclear reactors: crisis analysis, sextant, a general purpose physical analyser

    International Nuclear Information System (INIS)

    Barbet, N.; Dumas, M.; Mihelich, G.; Souchet, Y.; Thomas, J.B.

    1987-04-01

    Two developments of expert systems intended to work on line to the analysis of nuclear reactor transients are reported. During an hypothetical crisis occurring in a nuclear facility, a staff of the Institute for Protection and Nuclear Safety (IPSN) has to assess the risk to local population. The expert system is intended to work as an assistant to the staff. At the present time, it deals with the availability of the safety systems of the plant (e.g. ECCS), depending on the functional state of the support systems. A next step is to take into account the physical transient of the reactor (mass and energy balance, pressure, flows). In order to reach this goal as in the development of other similar expert systems, a physical analyser is required. This is the aim of SEXTANT, which combines several knowledge bases concerning measurements, models and qualitative behaviour of the plant with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. Both expert systems require powerful shells for their development. SPIRAL is such a toolkit for the development of expert systems devoted to the computer aided management of complex processes

  18. LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

    International Nuclear Information System (INIS)

    McElroy, W.N.

    1985-08-01

    This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old) was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence

  19. Development of a sixth-generation down-flow hanging sponge (DHS) reactor using rigid sponge media for post-treatment of UASB treating municipal sewage.

    Science.gov (United States)

    Onodera, Takashi; Tandukar, Madan; Sugiyana, Doni; Uemura, Shigeki; Ohashi, Akiyoshi; Harada, Hideki

    2014-01-01

    A sixth-generation down-flow hanging sponge reactor (DHS-G6), using rigid sponge media, was developed as a novel aerobic post-treatment unit for upflow anaerobic sludge blanket (UASB) treating municipal sewage. The rigid sponge media were manufactured by copolymerizing polyurethane with epoxy resin. The UASB and DHS system had a hydraulic retention time (HRT) of 10.6 h (8.6 h for UASB and 2 h for DHS) when operated at 10-28 °C. The system gave reasonable organic and nitrogen removal efficiencies. The final effluent had a total biochemical oxygen demand of only 12 mg/L and a total Kjeldahl nitrogen content of 6 mg/L. The DHS reactor gave particularly good nitrification performance, which was attributed to the new rigid sponge media. The sponge media helped to provide a sufficient HRT, and retained a high biomass concentration, extending the solids retention time. The DHS reactor maintained a high dissolved oxygen concentration under natural ventilation. Copyright © 2013 Elsevier Ltd. All rights reserved.

  20. REMOVAL OF ORGANIC MATTER AND TOXICITY IN AN UPFLOW IMMOBILIZED BIOMASS ANAEROBIC REACTOR TREATING HOSPITAL WASTEWATER: PRELIMINARY EVALUATION

    Directory of Open Access Journals (Sweden)

    MÓNICA PORRAS TORRES

    2013-01-01

    Full Text Available El objetivo de esta investigación consistió en evaluar el desempeño de un reactor anaerobio de flujo ascendente de biomasa inmovilizada (RAFABI tratando un efluente hospitalario real. Se estudió la remoción de materia orgánica y toxicidad, por medio de análisis como UV254, DQOfiltrada y determinación del porcentaje de inhibición en el crecimiento de la raíz de la cebolla. Los resultados mostraron que el proceso biológico estuvo estable durante los 287 días de operación continua, el valor medio de la relación AI/AP fue de 1.21±0.08, indicando que no hubo acumulación de ácidos en el sistema. Sin embargo, los valores de la eficiencia de remoción de DQOfiltrada, 56±15% y UV254, 21±36%, no fueron representativos. La toxicidad se redujo en 50%. Con base en lo anterior, es necesario utilizar el reactor anaerobio en combinación con otros procesos como por ejemplo los procesos de oxidación avanzada, para continuar reduciendo la materia orgánica recalcitrante al proceso anaerobio. Se comprobó la capacidad que tienen los reactores anaerobios de biomasa inmovilizada para remover la toxicidad.

  1. Experimental studies and mathematical modeling of an up-flow biofilm reactor treating mustard oil rich wastewater.

    Science.gov (United States)

    Chakraborty, Chandrima; Chowdhury, Ranjana; Bhattacharya, Pinaki

    2011-05-01

    Bioremediation of lipid-rich model wastewater was investigated in a packed bed biofilm reactor (anaerobic filter). A detailed study was conducted about the influence of fatty acid concentration on biomethanation of the high-fat liquid effluent of edible oil refineries. The biochemical methane potential (BMP) of the liquid waste was reported and maximum cumulative methane production at the exit of the reactor is estimated to be 785 ml CH(4) (STP)/(gVSS added). The effects of hydraulic retention time (HRT), organic loading rate (OLR) and bed porosity on the cold gas efficiency or energy efficiency of the bioconversion process were also investigated. Results revealed that the maximum cold gas efficiency of the process is 42% when the total organic load is 2.1 g COD/l at HRT of 3.33 days. Classical substrate uninhibited Monod model is used to generate the differential system equations which can predict the reactor behavior satisfactorily. Copyright © 2011 Elsevier Ltd. All rights reserved.

  2. Nuclear and Physical Properties of Dielectrics under Neutron Irradiation in Fast (BN-600) and Fusion (DEMO-S) Reactors

    Science.gov (United States)

    Blokhin, D. A.; Chernov, V. M.; Blokhin, A. I.

    2017-12-01

    Nuclear and physical properties (activation and transmutation of elements) of BN and Al2O3 dielectric materials subjected to neutron irradiation for up to 5 years in Russian fast (BN-600) and fusion (DEMO-S) reactors were calculated using the ACDAM-2.0 software complex for different post-irradiation cooling times (up to 10 years). Analytical relations were derived for the calculated quantities. The results may be used in the analysis of properties of irradiated dielectric materials and may help establish the rules for safe handling of these materials.

  3. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  4. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  5. Program MCU for Monte-Carlo calculations of neutron-physical characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Abagyan, L.P.; Alekseev, N.I.; Bryzgalov, V.I.; Glushkov, A.E.; Gomin, E.A.; Gurevich, M.I.; Kalugin, M.A.; Majorov, L.V.; Marin, S.V.; Yhdkevich, M.S.

    1994-01-01

    A description of the MCU data modification is presented. The calculation results by the MCU-2 and MCU-3 codes are compared for the critical assemblies of a different reactor types. The full list of the critical assemblies calculation results obtained by all MCU code versions is given. 32 refs.; 32 tabs

  6. Neutron physics and nuclear data measurements with accelerators and research reactors

    International Nuclear Information System (INIS)

    1985-08-01

    The report contains a collection of lectures devoted to the latest theoretical and experimental developments in the field of fast neutron measurements and in the studies of neutron interactions with nuclei. The possibilities offered by particle accelerators and research reactors for research and technological applications in these fields are pointed out

  7. Oklo: The fossil nuclear reactors. Physics study - Translation of chapters 6, 13 and conclusions

    International Nuclear Information System (INIS)

    Naudet, R.

    1996-09-01

    Three parts of the 1991 book 'Oklo: reacteurs nucleaires fossiles. Etude physique' have been translated in this report. The chapters bear the titles 'Study of criticality'(45 p.), 'Some problems with the overall functioning of the reactor zones'(45 p.) and 'Conclusions' (15 p.), respectively

  8. Oklo: The fossil nuclear reactors. Physics study - Translation of chapters 6, 13 and conclusions

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, R. [CEA, Paris (France)

    1996-09-01

    Three parts of the 1991 book `Oklo: reacteurs nucleaires fossiles. Etude physique` have been translated in this report. The chapters bear the titles `Study of criticality`(45 p.), `Some problems with the overall functioning of the reactor zones`(45 p.) and `Conclusions` (15 p.), respectively.

  9. Efficiency of different techniques of physical flattening by fuel while selection of optimum arrangement of large fast reactor core

    International Nuclear Information System (INIS)

    Grachev, E.A.; Dejnega, N.L.; Mitin, A.M.

    1974-01-01

    Results are given of calculations for selecting the parameters of the large fast breeder reactor core (1500 Mw) operating on oxide fuel with a sodium coolant. A complex optimum criterion was selected for energy intensity, energy distribution, breeding ratio and critical load factor, run duration, burning, reactivity variations, influence of CV3, fuel overloads, and calculated fue fuel expenses. The effectivities of various methods for physical grading of fuel (enrichment and composition) were examined in accordance with the optimum criterion. Parameters of reactor cores optimum arrangements are presented. Continuous reactor operation during 0.8-1.0 yr. at energy intensity more than 400 kW was shown to be essential for attaining the optimum chosen. Accounting for the CV3 system and partial fuel overloads, the methods of balancing energy release either by enriching fuel or by changing its composition proved to be almost equally effective. All calculations were performed with the aid of a 18-4-RZ-15B program on the basis of a BNAB-26 constant system [ru

  10. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2014-01-01

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM

  11. Strategy to identify the causes and to solve a sludge granulation problem in methanogenic reactors: application to a full-scale plant treating cheese wastewater.

    Science.gov (United States)

    Macarie, Hervé; Esquivel, Maricela; Laguna, Acela; Baron, Olivier; El Mamouni, Rachid; Guiot, Serge R; Monroy, Oscar

    2017-08-26

    Granulation of biomass is at the basis of the operation of the most successful anaerobic systems (UASB, EGSB and IC reactors) applied worldwide for wastewater treatment. Despite of decades of studies of the biomass granulation process, it is still not fully understood and controlled. "Degranulation/lack of granulation" is a problem that occurs sometimes in anaerobic systems resulting often in heavy loss of biomass and poor treatment efficiencies or even complete reactor failure. Such a problem occurred in Mexico in two full-scale UASB reactors treating cheese wastewater. A close follow-up of the plant was performed to try to identify the factors responsible for the phenomenon. Basically, the list of possible causes to a granulation problem that were investigated can be classified amongst nutritional, i.e. related to wastewater composition (e.g. deficiency or excess of macronutrients or micronutrients, too high COD proportion due to proteins or volatile fatty acids, high ammonium, sulphate or fat concentrations), operational (excessive loading rate, sub- or over-optimal water upflow velocity) and structural (poor hydraulic design of the plant). Despite of an intensive search, the causes of the granulation problems could not be identified. The present case remains however an example of the strategy that must be followed to identify these causes and could be used as a guide for plant operators or consultants who are confronted with a similar situation independently of the type of wastewater. According to a large literature based on successful experiments at lab scale, an attempt to artificially granulate the industrial reactor biomass through the dosage of a cationic polymer was also tested but equally failed. Instead of promoting granulation, the dosage caused a heavy sludge flotation. This shows that the scaling of such a procedure from lab to real scale cannot be advised right away unless its operability at such a scale can be demonstrated.

  12. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  13. Ecotoxicity evaluation of a WWTP effluent treated by solar photo-Fenton at neutral pH in a raceway pond reactor.

    Science.gov (United States)

    Freitas, A M; Rivas, G; Campos-Mañas, M C; Casas López, J L; Agüera, A; Sánchez Pérez, J A

    2017-01-01

    Some pollutants can be resistant to wastewater treatment, hence becoming a risk to aquatic and terrestrial biota even at the very low concentrations (ng L -1 -μg L -1 ) they are commonly found at. Tertiary treatments are used for micropollutant removal but little is known about the ecotoxicity of the treated effluent. In this study, a municipal secondary effluent was treated by a solar photo-Fenton reactor at initial neutral pH in a raceway pond reactor, and ecotoxicity was evaluated before and after micropollutant removal. Thirty-nine micropollutants were identified in the secondary effluent, mainly pharmaceuticals, with a total concentration of ≈80 μg L -1 . After treatment, 99 % microcontaminant degradation was reached. As for ecotoxicity reduction, the assayed organisms showed the following sensitivity levels: Tetrahymena thermophila > Daphnia magna > Lactuca sativa > Spirodela polyrhiza ≈ Vibrio fischeri. The initial effluent showed an inhibitory effect of 40 % for T. thermophila and 20 % for D. magna. After 20 min of photo-Fenton treatment, no toxic effect was observed for T. thermophila and toxicity dropped to 5 % for D. magna. Graphical abstract Ecotoxicity removal by solar photo-Fenton at neutral pH. ᅟ.

  14. Use of electrochemical oxidation process as post-treatment for the effluents of a UASB reactor treating cellulose pulp mill wastewater.

    Science.gov (United States)

    Buzzini, A P; Miwa, D W; Motheo, A J; Pires, E C

    2006-01-01

    The main purpose of this study was to evaluate the performance of the electrochemical oxidation process as a post-treatment for the effluents of a bench-scale UASB reactor treating simulated wastewater from an unbleached pulp plant. The oxidation process was performed using a single compartment cell with two plates as electrodes. The anode was made of Ti/Ru0.3Ti0.7O2 and the cathode of stainless steel. The following variables were evaluated: current density (75, 150 and 225 mA cm(-2)) and recirculation flow rate in the electrochemical cell (0.22, 0.45 and 0.90 L h(-1)). The increase in current density from 75 to 225 mA cm(-2) did not increased the color removal efficiency for the tested flow rates, 0.22, 0.45 and 0.90 L h(-1), however the energy consumption increased significantly. The results indicated the technical feasibility of the electrochemical treatment as post-treatment for UASB reactors treating wastewaters from pulp and paper plants.

  15. The effect and biological mechanism of COD/TN ratio on nitrogen removal in a novel upflow microaerobic sludge reactor treating manure-free piggery wastewater.

    Science.gov (United States)

    Li, Jianzheng; Meng, Jia; Li, Jiuling; Wang, Cheng; Deng, Kaiwen; Sun, Kai; Buelna, Gerardo

    2016-06-01

    A novel upflow microaerobic sludge reactor (UMSR) was constructed to treat manure-free piggery wastewater with high NH4(+)-N concentration and low COD/TN ratio, and the effect and biological mechanism of COD/TN ratio on nitrogen removal were investigated at a constant hydraulic retention time of 8h and 35°C. The results showed that the UMSR could treat the wastewater with a better synchronous removal of COD, NH4(+)-N and TN. The microaerobic UMSR allowed nitrifiers, and heterotrophic and autotrophic denitrifiers to thrive in the flocs, revealing a multiple nitrogen removal mechanism in the reactor. Both the nitrifiers and denitrifiers would be restricted by an influent COD/TN ratio more than 0.82, resulting in a decrease of TN removal in the UMSR. To get a TN removal over 80% with a TN load removal above 0.86kg/(m(3)·d) in the UMSR, the influent COD/TN ratio should be less than 0.70. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    Science.gov (United States)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  17. Coupled multi-physics simulation frameworks for reactor simulation: A bottom-up approach

    International Nuclear Information System (INIS)

    Tautges, Timothy J.; Caceres, Alvaro; Jain, Rajeev; Kim, Hong-Jun; Kraftcheck, Jason A.; Smith, Brandon M.

    2011-01-01

    A 'bottom-up' approach to multi-physics frameworks is described, where first common interfaces to simulation data are developed, then existing physics modules are adapted to communicate through those interfaces. Physics modules read and write data through those common interfaces, which also provide access to common simulation services like parallel IO, mesh partitioning, etc.. Multi-physics codes are assembled as a combination of physics modules, services, interface implementations, and driver code which coordinates calling these various pieces. Examples of various physics modules and services connected to this framework are given. (author)

  18. In-core fuel management for the course on operational physics of power reactors

    International Nuclear Information System (INIS)

    Levine, S.H.

    1982-01-01

    The heart of a nuclear power station is the reactor core producing power from the fissioning of uranium or plutonium fuel. Expertise in many different technical fields is required to provide fuel for continuous economical operation of a nuclear power plant. In general, these various technical disciplines can be dichotomized into ''Out-of-core'' and ''In-core'' fuel management. In-core fuel management is concerned, as the name implies, with the reactor core itself. It entails calculating the core reactivity, power distribution, and isotopic inventory for the first and subsequent cores of a nuclear power plant to maintain adequate safety margins and operating lifetime for each core. In addition, the selection of reloading schemes is made to minimize energy costs

  19. Status and future program of reactor physics experiments in JAERI Critical facilities, FCA and TCA

    International Nuclear Information System (INIS)

    Okajima, Shigeaki; Osugi, Toshitaka; Nakajima, Ken; Suzaki, Takenori; Miyoshi, Yoshinori

    1999-01-01

    The critical facilities in JAERI, FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly), have been used to provide integral data for evaluation of nuclear data as well as for development of various types of reactor since they went critical in 1960's. In this paper a review is presented on the experimental programs in both facilities. And the experimental programs in next 5 years are also shown. (author)

  20. Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

    International Nuclear Information System (INIS)

    Olson, W.O.; Nigg, D.W.

    1981-05-01

    The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft