WorldWideScience

Sample records for reactor physics characteristics

  1. 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

  2. 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

  3. Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

  4. 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

  5. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    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

  6. 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.

  7. 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.

  8. 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

  9. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  10. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  11. Physical characteristics of non-fuel assembly reactor components

    International Nuclear Information System (INIS)

    Hawkes, E.C.

    1994-09-01

    The primary objective of this report is to enhance the utility of the Characteristics Data Base (CDB). This has been accomplished by providing a pictorial representation of the principal non-fuel assembly (NFA) components along with a tabular summary of key information about each type of component. This report is intended for use as an adjunct to the CDB. Toward this end, the report may be used either as a complement to the detailed descriptions in the CDB, or as a stand-alone document that acts as an illustrated abstract of the CDB. Line drawings of major NFA components are included. Data not provided in the CDB are also included. Summary descriptions of each component are given in tabular format

  12. 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)

  13. 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)

  14. Experimental investigation of the neutron physics characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang; Thong, Ha Van [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    The investigation of the neutron physics characteristics of the Dalat Reactor has obtained the results as follows: 1/ The effective fraction of delayed photoneutrons and the extraneous neutron source left after reactor shut down are measured. 2/ The lowest power levels of critical states of the reactor are determined. 3/The perturbation effect is investigated when a water column or a plexiglass rod is substituted for a fuel element. 4/ The relative axial and radial distributions of the thermal neutrons measured and the geometrical parameters of the core such as the inhomogeneous coefficients, the buckling, the effective height and radius, the extrapolated distances are obtained. 4/ The thermal neutron distributions are measured around the old graphite reflector. (author). 10 refs., 10 figs., 2 tabs.

  15. 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

  16. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

    These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

  17. 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

  18. 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

  19. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  20. NF-6 program complex for BESM-6 computation of the basic neutron-physical characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Zizin, M.N.; Savochkina, O.A.; Chukhlova, O.P.

    1978-01-01

    A structure of standard designations is described and semantics of a number of standard values used in a NF-6 program complex is given. Main source data and results of neutron-physical reactor calculation are standard values, the peculiarities of FORTRAN and ALGOL-GDR algorithm languages in the DUBNA monitoring system were taken account of. As a base of standard values list the FIHAR system list, supplemented with new standard designations for integral reactor characteristics, is used. Developed is also a list of standard values to organize the exchange with external memory in the process of task solution and long-range storage

  1. Impact of the 37M fuel design on reactor physics characteristics

    International Nuclear Information System (INIS)

    Perez, R.; Ta, P.

    2013-01-01

    For CANDU nuclear reactors, aging of the Heat Transport System (HTS) leads to, among other effects, a reduction on the Critical Heat Flux (CHF) and dryout margin. In an effort to mitigate the impact of aging of the HTS on safety margins, Bruce Power is introducing a design change to the standard 37-element fuel bundle known as the modified 37-element fuel bundle, or 37M for short. As part of the overall design change process it was necessary to assess the impact of the modified fuel bundle design on key reactor physics parameters. Quantification of this impact on lattice cell properties, core reactivity properties, etc., was reached through a series of calculations using state-of-the-art lattice and core physics models, and comparisons against results for the standard fuel bundle. (author)

  2. 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

  3. 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.

  4. 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

  5. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  6. 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

  7. 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

  8. 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

  9. Neutron-physical characteristics of the TVRM-100 reactor with ten ring fuel channels

    International Nuclear Information System (INIS)

    Mikhajlov, V.M.; Myrtsymova, L.A.

    1988-01-01

    Three-dimensional heterogeneous calculations of TVRM-100 reactor which is a research reactor using enriched fuel with heavy-water moderator, coolant and reflector, are conducted. Achievable burnup depths depending on the number of removable FAs are presented. The maximum non-perturbed thermal neutron flux in the reflector is (2-1.8)x10 15 cm -2 c -1 ; mean flux on the fuel is 2.9x10 14 cm -2 c -1 . Energy release radial non-uniformity is 0.67, maximum bending by FA is ∼3.7. Reactivity temperature effect is negative and is equal to - 0.9x10 -4 grad -1 without accounting for experimental channels. Control rod efficiency in the radial reflector is high, but their location dose to experimental devices in the high neutron flux area is undesirable. 4 refs.; 5 figs

  10. Physical Characteristics of the Dalat Nuclear Research Reactor; Cac dac trung vat ly lo cua lo phan ung hat nhan Da Lat

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [ed.; Nuclear Research Inst., Da Lat (Viet Nam)

    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.

  11. The analysis of neutron physical characteristics of fast reactors by means of pulsed experiments

    International Nuclear Information System (INIS)

    Stumbur, Eh.A.; Milyutina, Z.N.

    1992-01-01

    Possibility is considered for determination of macroscopic cross sections of homogeneous multiplicating media with fast neutrons. It is shown that by means of the critical size, laplaccian and neutron pulse damping decrement measurement results it is possible to obtain values of almost all cross sections of a medium. The method is tested with systems of metal 235 U and BFS-32 assemblies with the composition, typical for fast power reactors. A suitable algorithm is developed for solving nonstationary asymptotic transport problems. Calculation results are compared with experimental ones. 21 refs.; 2 figs.; 3 tabs

  12. Identification of nuclear reactor characteristics by the reactor noise analysis

    International Nuclear Information System (INIS)

    Yashima, Hideyuki

    1980-01-01

    Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

  13. Practice of calculation of neutron-physical characteristics of reactors and radiating shielding in structure SNPS with program complex MCNP

    International Nuclear Information System (INIS)

    Krotov, A.D.; Son'ko, A.V.

    2009-01-01

    Calculation of neutron-physical properties and radiation protection of space power reactor was made by means of the MCNP code allowing simulation of neutron, γ- and electron transport by the Monte Carlo method in the systems with combined geometry. Universality of the MCNP code has been demonstrated both for the calculation of reactor-converter so for the optimization of radiation protection that allows to reserve a new level of complex simulation of SNPS [ru

  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 problems on HCPWR

    International Nuclear Information System (INIS)

    Ishiguro, Yukio; Akie, Hiroshi; Kaneko, Kunio; Sasaki, Makoto.

    1986-01-01

    Reactor physics problems on high conversion pressurized water reactors (HCPWRs) are discussed. Described in this report are outline of the HCPWR, expected accuracy for the various reactor physical qualities, and method for K-effective calculation in the resonance energy area. And requested further research problems are shown. The target value of the conversion ratio are also discussed. (author)

  17. 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

  18. Research on reactor physics data

    International Nuclear Information System (INIS)

    1961-01-01

    In the early years of nuclear reactor research, each national program tended to develop its own reactor physics information. The Government of Norway proposed to the Agency the undertaking of a joint program in reactor physics utilizing the facilities and staff of its zero power reactor NORA then under construction. Following the approval by the Board of Governors in February, the Agency invited Member States to submit the names and qualifications of scientists they wished to suggest for the project. All the results and information gained through the program, which is expected to last about three years, will be placed at the disposal of the Agency's Member States

  19. 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.)

  20. 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)

  1. 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

  2. IRT-type research reactor physical calculation methodology

    International Nuclear Information System (INIS)

    Carrera, W.; Castaneda, S.; Garcia, F.; Garcia, L.; Reyes, O.

    1990-01-01

    In the present paper an established physical calculation procedure for the research reactor of the Nuclear Research Center (CIN) is described. The results obtained by the method are compared with the ones reported during the physical start up of a reactor with similar characteristics to the CIN reactor. 11 refs

  3. Physical model of reactor pulse

    International Nuclear Information System (INIS)

    Petrovic, A.; Ravnik, M.

    2004-01-01

    Pulse experiments have been performed at J. Stefan Institute TRIGA reactor since 1991. In total, more than 130 pulses have been performed. Extensive experimental information on the pulse physical characteristics has been accumulated. Fuchs-Hansen adiabatic model has been used for predicting and analysing the pulse parameters. The model is based on point kinetics equation, neglecting the delayed neutrons and assuming constant inserted reactivity in form of step function. Deficiencies of the Fuchs-Hansen model and systematic experimental errors have been observed and analysed. Recently, the pulse model was improved by including the delayed neutrons and time dependence of inserted reactivity. The results explain the observed non-linearity of the pulse energy for high pulses due to finite time of pulse rod withdrawal and the contribution of the delayed neutrons after the prompt part of the pulse. The results of the improved model are in good agreement with experimental results. (author)

  4. Development of physical conceptions of fast reactors

    International Nuclear Information System (INIS)

    Khomyakov, Yu.S.; Matveev, V.I.; Moiseev, A.V.

    2013-01-01

    • Russian experience in developing fast reactors has proved clearly scientific justification of conceptual physical principles and their technical feasibility. • However, the potential of fast reactors caused by their physical features has not been fully realized. • In order to assure the real possibility of transition to the nuclear power with fast reactors by about 2030 it is necessary to consistently update fast reactor designs for solving the following key problems: - increasing of self-protection level of reactor core; - improvement of technical and economical characteristics; - solution of the problems related to the fuel supply of nuclear power and assimilation of closed nuclear fuel cycle; - disposal of long lived radioactive waste and transmutation of minor actinides. • Russian program (2010-2020) on the development of basic concepts of the new generation reactors implies successive solution of the above problems. • New technical decisions will be demonstrated by development and assimilation of the new reactors: - BN-800 – development of the fuel cycle infrastructure and mastering of the new types of fuel; - BN-1200 reactor – demonstration economical efficiency of fast reactor and new level of safety; - BREST development and demonstration new heavy liquid metal coolant technology and alternative design concept

  5. Effects of fuel enrichment on the physics characteristics of plutonium-fueled light water high converter reactors

    International Nuclear Information System (INIS)

    Chawla, R.; Seiler, R.; Gmur, K.

    1986-01-01

    Investigations have been carried out for three additional cores of the phase 1 experimental program on light water high converter reactor test lattices in the PROTEUS facility. An 8% (average) fissile plutonium tight-pitch lattice with a fuel/moderator volumetric ratio of 2.0 was considered. As for the earlier reported 6% (average) fissile plutonium test lattice, H 2 O, Dowtherm, and air were the moderator state investigated. Significant enrichment-dependent trends have been identified in the comparisons of calculated and experimental results for the wet (moderated cases, particularly for the important reaction rate ratio of 238 U capture of 239 Pu fission. These are then reflected in the comparison of moderator voidage characteristics, expressed in terms of individual components of the kinfinity void coefficient

  6. Effects of fuel enrichment on the physics characteristics of plutonium-fueled light water high converter reactors

    International Nuclear Information System (INIS)

    Chawla, R.; Seiler, R.; Gmuer, K.

    1986-01-01

    Investigations have been carried out for three additional cores of the phase 1 experimental program on light water high converter reactor test lattices in the PROTEUS facility. An 8% (average) fissile plutonium tight-pitch lattice with a fuel/moderator volumetric ratio of 2.0 was considered. As for the earlier reported 6% (average) fissile plutonium test lattice, H 2 O, Dowtherm, and air were the moderator states investigated. Significant enrichment-dependent trends have been identified in the comparisons of calculated and experimental results for the wet (moderated) cases, particularly for the important reaction rate ratio of 238 U capture to 239 Pu fission. These are then reflected in the comparison of moderator voidage characteristics, expressed in terms of individual components of the k-infinity void coefficient. (author)

  7. Fast reactor physics - an overview

    International Nuclear Information System (INIS)

    Lee, S.M.

    2004-01-01

    An introduction to the basic features of fast neutron reactors is made, highlighting the differences from the more conventional thermal neutron reactors. A discussion of important feedback reactivity mechanisms is given. Then an overview is presented of the methods of fast reactor physics, which play an important role in the successful design and operation of fast reactors. The methods are based on three main elements, namely (i) nuclear data bases, (ii) numerical methods and computer codes, and (iii) critical experiments. These elements are reviewed and the present status and future trends are summarized. (author)

  8. 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)

  9. 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

  10. Reactors and physics education

    International Nuclear Information System (INIS)

    Hayter, J.B.

    1992-01-01

    This paper discussed some ideas for using neutrons in physics education, including experiments which demonstrate diffraction and optical refraction, divergence imaging, Zeeman splitting, polarization, Larmor precession, and neutron spin-echo. (author)

  11. 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

  12. TRIGA reactor health physics considerations

    International Nuclear Information System (INIS)

    Johnson, A.G.

    1970-01-01

    The factors influencing the complexity of a TRIGA health physics program are discussed in details in order to serve as a basis for later consideration of various specific aspects of a typical TRIGA health physics program. The health physics program must be able to provide adequate assistance, control, and safety for individuals ranging from the inexperienced student to the experienced postgraduate researcher. Some of the major aspects discussed are: effluent release and control; reactor area air monitoring; area monitoring; adjacent facilities monitoring; portable instrumentation, personnel monitoring. TRIGA reactors have not been associated with many significant occurrences in the area of health physics, although some operational occurrences have had health physics implications. One specific occurrence at OSU is described involving the detection of non-fission-product radioactive particulates by the continuous air monitor on the reactor top. The studies of this particular situation indicate that most of the particulate activity is coming from the rotating rack and exhausting to the reactor top through the rotating rack loading tube

  13. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

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

    2000-01-01

    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.)

  14. Operational power reactor health physics

    International Nuclear Information System (INIS)

    Watson, B.A.

    1987-01-01

    Operational Health Physics can be comprised of a multitude of organizations, both corporate and at the plant sites. The following discussion centers around Baltimore Gas and Electric's (BG and E) Calvert Cliffs Nuclear Power Plant, located in Lusby, Maryland. Calvert Cliffs is a twin Combustion Engineering 825 MWe pressurized water reactor site with Unit I having a General electric turbine-generator and Unit II having a Westinghouse turbine-generator. Having just completed each Unit's ten-year Inservice Inspection and Refueling Outge, a total of 20 reactor years operating health physics experience have been accumulated at Calvert Cliffs. Because BG and E has only one nuclear site most health physics functions are performed at the plant site. This is also true for the other BG and E nuclear related organizations, such as Engineering and Quality Assurance. Utilities with multiple plant sites have corporate health physics entity usually providing oversight to the various plant programs

  15. Nuclear characteristic simulation device for reactor core

    International Nuclear Information System (INIS)

    Arakawa, Akio; Kobayashi, Yuji.

    1994-01-01

    In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

  16. A study on physical characteristics of supercritical light - water reactor loaded with (232U-238Th-238U) oxide fuel

    International Nuclear Information System (INIS)

    Kulikov, E. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, G. G.

    2007-01-01

    The attractiveness of using (U-Th)-fuel in supercritical light water reactor is considered. The dilution of 2 33U in 2 38U is proposed with the purpose of increasing non-proliferation of this fissile isotope. Comparison of different fuel compositions is accomplished from the point of view of fissile isotope breeding and achieved burn-up; parasitic neutron absorption cross-sections are also compared. It is analyzed the impact for neutron balance of both cladding materials: zirconium alloy and stainless steel

  17. WWER reactor physics code applications

    International Nuclear Information System (INIS)

    Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.

    1994-01-01

    The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs

  18. 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)

  19. 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

  20. 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

  1. 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

  2. Characteristics of HTTR's startup physics tests

    International Nuclear Information System (INIS)

    Nojiri, N.; Nakano, M.; Takeuchi, M.; Pohl, P.; Yamashita, K.

    1997-01-01

    The High Temperature Engineering Test Reactor (HTTR) which is under construction by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium gas-cooled reactor with an outlet temperature of 950 deg. C and a thermal output of 30MW. The first criticality is expected at the end of October 1997. The start-up physics tests (SPTs) are planned in the period from mid 1997 to the end of 1998. Characteristic items of the SPTs are: 1) Criticality approach; 2) Tests on a preliminary annual core; 3) Measurement of scram reactivity; 4) Excess reactivity test; 5) Measurements along with a 2-step-scram reactor shutdown procedure. (author)

  3. Physics and kinetics of TRIGA reactor

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This training module is written as an introduction to reactor physics for reactor operators. It assumes the reader has a basic, fundamental knowledge of physics, materials and mathematics. The objective is to provide enough reactor theory knowledge to safely operate a typical research reactor. At this level, it does not necessarily provide enough information to evaluate the safety aspects of experiment or non-standard operation reviews. The material provides a survey of basic reactor physics and kinetics of TRIGA type reactors. Subjects such as the multiplication factor, reactivity, temperature coefficients, poisoning, delayed neutrons and criticality are discussed in such a manner that even someone not familiar with reactor physics and kinetics can easily follow. A minimum of equations are used and several tables and graphs illustrate the text. (author)

  4. Reactor physics challenges in GEN-IV reactor design

    International Nuclear Information System (INIS)

    Driscoll, Michael K.; Hejzlar, Pavel

    2005-01-01

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

  5. 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

  6. Reactor physics challenges in GEN-IV reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

    2005-02-15

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

  7. Inherent safety characteristics of innovative reactors

    International Nuclear Information System (INIS)

    Heil, J.A.

    1995-11-01

    The added safety value of innovative or third generation reactor designs has been evaluated in order to determine the most suitable candidate for Dutch government funded research and development support. To this end, four innovative reactor concepts, viz. PIUS (Process Inherent Ultimate Safety), PRISM (Power Reactor Innovative Small), HTR-M (High Temperature Reactor Module) and MHTGR (Modular High Temperature Gas-cooled Reactor), have been studied and their passive and inherent safety characteristics have been outlined. Also the outlook for further technological and industrial development has been considered. The results of the study confirm the perspective of the innovative reactors for reduced dependence on active safety provisions and for a further reduced vulnerability to technical failures and human errors. The accident responses to generic accident initiators, viz. reactivity and cooling accidents, and also to reactor specific accidents show that neither active safety systems nor short term operator actions are required for maintaining the reactor core in a controlled and coolable condition. Whether this gives rise to a higher total safety of the innovative reactor designs, compared to evolutionary or advanced reactors, cannot be concluded. Supplementary experimental and analytical analyses of reactor specific accidents are required to be able to assess the safety of these innovative designs in a more quantitative manner. It is believed that the safety case of innovative reactors, which are less dependent on active safety systems, can be communicated with the general public in a more transparent way. Considering the perspective for further technological and industrial development it is not expected that any of the considered innovative reactor concepts will become commercially available within the next one to two decades. However, they could be made available earlier if they would receive sufficient financial backing. Considering the added safety perspectives

  8. Opportunities for reactor scale experimental physics

    International Nuclear Information System (INIS)

    1999-01-01

    A reactor scale tokamak plasma will exhibit three areas of physics phenomenology not accessible by contemporary experimental facilities. These are: (1) instabilities generated by energetic alpha particles; (2) self-heating phenomena; and (3) reactor scale physics, which includes integration of diverse physics phenomena, each with its own scaling properties. In each area, selected examples are presented that demonstrate the importance and uniqueness of physics results from reactor scale facilities for both inductive and steady state reactor options. It is concluded that the physics learned in such investigations will be original physics not attainable with contemporary facilities. In principle, a reactor scale facility could have a good measure of flexibility to optimize the tokamak approach to magnetic fusion energy. (author)

  9. 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)

  10. Activity report of Reactor Physics Division - 1993

    International Nuclear Information System (INIS)

    Indira, R.

    1994-01-01

    The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1993 are reported. The activities are arranged under the headings: Nuclear Data Processing and validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. (author). refs., figs., tabs

  11. Activity report of Reactor Physics Division-1995

    International Nuclear Information System (INIS)

    Gopalakrishnan, V.

    1996-01-01

    The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1995 are reported. The activity are arranged under the headings: Nuclear Data Processing and Validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. refs., figs., tabs

  12. Activity report of Reactor Physics Division - 1993

    Energy Technology Data Exchange (ETDEWEB)

    Indira, R [ed.; Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1994-12-31

    The research and development (R and D) activities of the Reactor Physics Division of Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1993 are reported. The activities are arranged under the headings: Nuclear Data Processing and validation, Core Physics and Operation Studies, Reactor Kinetics and Safety analysis, Reactor Noise Analysis and Radiation Transport and Shielding Studies. List of publication is given at the end. (author). refs., figs., tabs.

  13. Reactor physics methods development at Westinghouse

    International Nuclear Information System (INIS)

    Mueller, E.; Mayhue, L.; Zhang, B.

    2007-01-01

    The current state of reactor physics methods development at Westinghouse is discussed. The focus is on the methods that have been or are under development within the NEXUS project which was launched a few years ago. The aim of this project is to merge and modernize the methods employed in the PWR and BWR steady-state reactor physics codes of Westinghouse. (author)

  14. Research reactor RB, technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.

    1978-01-01

    Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) [sr

  15. 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

  16. 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 �...

  17. Characteristics of a reactor with power reactivity feedback

    International Nuclear Information System (INIS)

    Li Fengyu; Zhang Yusheng; Zhang Guangfu; Liu Ying

    2008-01-01

    The point-reactor model with power reactivity feedback becomes a nonlinear system. Its dynamic characteristic shows great complexity. According to the mathematic definition of stability in differential equation qualitative theory, the model of a reactor with power reactivity feedback is judged unstable. The equilibrium point is a saddle-node point. A portion of the trajectory in the neighborhood of the equilibrium point is parabolic fan curve, and the other is hyperbolic fan curve. Based on phase locus near the equilibrium point, it is pointed out that the model is still stable within physical limits. The difference between stabilities in the mathematical sense and in the physical sense is indicated. (authors)

  18. Activity report of Reactor Physics Division - 1988

    International Nuclear Information System (INIS)

    Keshavamurthy, R.S.

    1989-01-01

    This report highlights the progress of activities carried out during the year 1988 in Reactor Physics Division in the form of brief summaries. The topics are organised under the following subject categories:(1) nuclear data evaluation , processing and validation, (2) core physics and analysis, (3) reactor kinetics and safety analysis, (4) noise analysis and (5) radiation transport and shielding. List of publications by the members of the Division and the Reactor Physics Seminars held during the year 1988, is included at the end of report. (author). refs., figs., tabs

  19. 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)

  20. 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

  1. Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

    Energy Technology Data Exchange (ETDEWEB)

    Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-09-01

    An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

  2. 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

  3. 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

  4. 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

  5. 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.)

  6. Physics of high-temperature reactors

    International Nuclear Information System (INIS)

    Massimo, L.

    1976-01-01

    The subject is covered in chapters entitled: general description of the HTR core; general considerations about reactor physics; neutron cross-sections; basic aspects of transport and diffusion theory; methods for the solution of the diffusion equation; slowing-down and thermalization in graphite; resonance absorption; spectrum calculations and cross-section averaging; burn-up; core design; fuel management and cost calculations; temperature coefficient; core dynamics and accident analysis; reactor control; peculiarities of HTR physics; analysis of calculational accuracy; sequence of reactor design calculations. (U.K.)

  7. 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)

  8. Summary of the progress of reactor physics in Japan reviewing the activities related to NEA Committee on Reactor Physics

    International Nuclear Information System (INIS)

    Hirota, Jitsuya

    1984-09-01

    The progress of fast and thermal reactor physics, fusion neutronics and shielding researches in these twenty years can be clearly recognized in the reviews of reactor physics activities in Japan which had been perpared by the Special Committee on Reactor Physics: the joint committee under Atomic Energy Society of Japan and JAERI. Many topics of those discussed at the NEACRP meetings concerned fast reactor physics. Information exchange on the topics such as adjustment of group cross sections by integral data, central worth discrepancy, sodium void effect and heterogeneous core stimulated the researches in Japan. And achievements in Japan including those in the JAERI Fast Critical Facility FCA were reported and contributed largely to the international co-operation. In addition, the contribution from Japan was also made concerning a study of fusion blanket. Among various specialists' meetings recommended by NEACRP, those on nuclear data and benchmarks for reactor shielding were often held since 1973 and helpful to the progress of shielding researches in Japan. The Third Specialists' Meeting on Reactor Noise (SMORN-III) was held in Tokyo in 1981, indicating the recent progress in safety-related applications of reactor noise analysis. The NEACRP benchmark tests were quite useful to the progress of reactor physics in Japan, which included the benchmark calculations of BWR lattice cell, key parameters and burn-up characteristics of a large LMFBR, FBR and PWR shielding, and so on. It may be noted that the benchmark test on reactor noise analysis methods was successfully conducted by Japan in connection with SMORN-III. In addition, the co-operation was positively made to the compilation of light water lattice data, and the preparation of reviews on actinide production and burn-up, and blanket physics. (J.P.N.)

  9. Inherently safe characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    1989-01-01

    This report is based on a detailed study which was carried out by Colenco (a company of the Motor-Columbus Group) on behalf of the Commission of the European Communities (CEC). It presents a summary of this study and concentrates more on the generic issues involved in the subject of inherent safety in nuclear power plants. It is assumed that the reader is reasonably familiar with the design outline of the systems included in the report. The report examines the role of inherent design features in achieving the safety of nuclear power plants as an alternative to the practice, which is largely followed in current reactors, of achieving safety by the addition of engineered safety features. The report examines current reactor systems to identify the extent to which their characteristics are either already inherently safe or, on the other hand, have inherent characteristics that require protective action to be taken. It then considers the advantages of introducing design changes to improve their inherent safety characteristics. Next, it looks at some new reactor types for which claims of inherent safety are made to see to what extent these claims are justified. The general question is then considered whether adoption of the inherently safe reactors would give advantages (by reducing risk in real terms or by improving the public acceptability of nuclear power) which are sufficient to offset the expected high costs and the technical risks associated with any new technology

  10. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

  11. Activity report of Reactor Physics Division : 1990

    International Nuclear Information System (INIS)

    Mohanakrishnan, P.

    1991-01-01

    The major Research and Development and Project activities carried out during the year 1990 in Reactor Physics Division are presented in the form of summaries in this report. The various activities are organised under the following areas : (1) Nuclear Data Evaluation, Processing and Validation, (2) Core Physics and Analysis, (3) Reactor Kinetics and Safety Analysis, (4) Noise Analysis, and (5) Radiation Transport and Shielding. FBTR was restarted in July 1990 and the power was raised upto 500 kW. A number of low power physics experiments on reactivity coefficients, kinetics and noise, neutron flux and gamma dose in B cells, were performed, which are discussed in this report. (author). figs., tabs

  12. 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

  13. 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

  14. 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)

  15. Activity Report of Reactor Physics Division - 1997

    International Nuclear Information System (INIS)

    Singh, Om Pal

    1998-01-01

    The research and development activities of the Reactor Physics Division of the Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam during 1997 are reported. The activities are arranged under the headings: nuclear data processing and validation, PFBR and KAMINI core physics, FBTR core physics, radioactivity and shielding and safety analysis. A list of publications of the Division and seminars delivered are included at the end of the report

  16. Compilation of reactor-physical data of the AVR experimental reactor for 1982

    International Nuclear Information System (INIS)

    Werner, H.; Wawrzik, U.; Grotkamp, T.; Buettgen, I.

    1983-12-01

    Since the end of 1981 the calculation model AVR-80 has been taken as a basis for compiling reactor-physical data of the AVR experimental reactor. A brief outline of the operation history of 1982 is given, including the beginning of a large-scale experiment dealing with change-over from high enriched uranium to low enriched uranium. Calculations relative to spectral shift, diffusion, temperature, burnup, and recirculation of the fuel elements are described in brief. The essential results of neutron-physical and thermodynamic calculations and the characteristical data of the various types of fuel used are shown in tables and illustrations. (RF) [de

  17. 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

  18. SERPENT Monte Carlo reactor physics code

    International Nuclear Information System (INIS)

    Leppaenen, J.

    2010-01-01

    SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)

  19. Activity report of Reactor Physics Division - 1989

    International Nuclear Information System (INIS)

    1990-01-01

    The highlights of the various studies carried out during the year 1989 in Reactor Physics Division are presented in this report in the form of summaries. The topics are organised under the following subjects: (1) nuclear data evaluation, processing and validation, (2) core physics and analysis, (3) reacto r kinetics and safety analysis, (4) noise analysis, and radiation transport and shielding. It is observed that with the restart and operation of FBTR at low power for some time, some of the low power physics experiments were completed and plans and procedures for the remaining physics experiments at intermediate and high power (upto 10 MWt) have been prepared. The lists of publications by the members of Division and the Reactor Physics Seminars held during the year 19 89, are included at the end of the report. (author). refs., figs., tabs

  20. Oscillation characteristics of the reactor 'A'; Oscilatorne karakteristike reaktora 'A'

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V; Lolic, B [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1961-07-01

    In addition to good knowledge of reactor physical properties, design of the reactor oscillator demands determining of the oscillator operating points as well as oscillation reactor properties. This paper contains study of the RA reactor power changes due to oscillations in in one of the vertical experimental channels. It has been concluded that the reactor optimum operating conditions are attained when the oscillator operates at optimum points, and other parameters are determined dependent on the sensitivity of the method and reactor stability.

  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 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)

  3. New trends in reactor physics design methods

    International Nuclear Information System (INIS)

    Jagannathan, V.

    1993-01-01

    Reactor physics design methods are aimed at safe and efficient management of nuclear materials in a reactor core. The design methodologies require a high level of integration of different calculational modules of many a key areas like neutronics, thermal hydraulics, radiation transport etc in order to follow different 3-D phenomena under normal and transient operating conditions. The evolution of computer hardware technology is far more rapid than the software development and has rendered such integration a meaningful and realizable proposition. The aim of this paper is to assess the state of art of the physics design codes used in Indian thermal power reactor applications with respect to meeting the design, operational and safety requirements. (author). 50 refs

  4. 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

  5. Technical specifications: Health Physics Research Reactor

    International Nuclear Information System (INIS)

    1979-02-01

    The 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. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

  6. Physical measurements in Marcoule reactors (1962)

    International Nuclear Information System (INIS)

    Teste du Bailler, A.

    1962-01-01

    A brief description of the physical measurements in Marcoule reactors is given here. During commissioning and subsequent years of operation, various experiments ha been carried out to check design data, and improve the operating conditions and also test theoretical models for kinetic studies. (author) [fr

  7. Reactor physics calculations on HTR type configurations

    Energy Technology Data Exchange (ETDEWEB)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.).

  8. Reactor physics calculations on HTR type configurations

    International Nuclear Information System (INIS)

    Klippel, H.T.; Hogenbirk, A.; Stad, R.C.L. van der; Janssen, A.J.; Kuijper, J.C.; Levin, P.

    1995-04-01

    In this paper a short description of the ECN nuclear analysis code system is given with respect to application in HTR reactor physics calculations. First results of calculations performed on the PROTEUS benchmark are shown. Also first results of a HTGR benchmark are given. (orig.)

  9. Safety characteristics of small heat producing reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1987-10-01

    The primary objectives of protection in nuclear power plants are the possibility to shut the reactor down in case of emergency and keep it subcritical in the long run, the existence of a heat sink for post-decay heat removal in order to avoid overheating, let alone core meltdown, and the containment of radioactivity within the barriers designed for this purpose, thus preventing significant activity release. In principle, these objectives can be met in various ways, namely by active, passive or inherent technical safeguards systems. In practice, a mixture of these approaches is employed in almost all cases. What matters in the end is the assessment of the overall concept, not of some outstanding feature. Inherent characteristics are easier to achieve in small reactors. However, also in this case, inherent safety does not mean absolute safety. If inherent safety characteristics were all encompassing, they would have to include self-healing effects. However, inanimate matter is incapable of such self-organization. Consequently, inherent characteristics in nuclear technology by definition should include the increased use of dissipative processes in the thermal part of the plant. (author)

  10. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  11. Ad hoc committee on reactor physics benchmarks

    International Nuclear Information System (INIS)

    Diamond, D.J.; Mosteller, R.D.; Gehin, J.C.

    1996-01-01

    In the spring of 1994, an ad hoc committee on reactor physics benchmarks was formed under the leadership of two American Nuclear Society (ANS) organizations. The ANS-19 Standards Subcommittee of the Reactor Physics Division and the Computational Benchmark Problem Committee of the Mathematics and Computation Division had both seen a need for additional benchmarks to help validate computer codes used for light water reactor (LWR) neutronics calculations. Although individual organizations had employed various means to validate the reactor physics methods that they used for fuel management, operations, and safety, additional work in code development and refinement is under way, and to increase accuracy, there is a need for a corresponding increase in validation. Both organizations thought that there was a need to promulgate benchmarks based on measured data to supplement the LWR computational benchmarks that have been published in the past. By having an organized benchmark activity, the participants also gain by being able to discuss their problems and achievements with others traveling the same route

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

    International Nuclear Information System (INIS)

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    experimental series that were performed at 17 different reactor facilities. The Handbook is organized in a manner that allows easy inclusion of additional evaluations, as they become available. Additional evaluations are in progress and will be added to the handbook periodically. Content: FUND - Fundamental; GCR - Gas Cooled (Thermal) Reactor; HWR - Heavy Water Moderated Reactor; LMFR - Liquid Metal Fast Reactor; LWR - Light Water Moderated Reactor; PWR - Pressurized Water Reactor; VVER - VVER Reactor; Evaluations published as drafts 2 - Related Information: International Criticality Safety Benchmark Evaluation Project (ICSBEP); IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments; IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan ; IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database ; IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility; IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation ; IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility ; IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation ; IRPHE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents; IRPHE-ARCH-01, Archive of HTR Primary Documents ; IRPHE/AVR, AVR High Temperature Reactor Experience, Archival Documentation ; IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters; IRPhE/BERENICE, effective delayed neutron fraction measurements ; IRPhE-TAPIRO-ARCHIVE, fast neutron source reactor primary documents, reactor physics experiments. The International Handbook of Evaluated Reactor Physics Benchmark Experiments was prepared by a working party comprised of experienced reactor physics personnel from Belgium, Brazil, Canada, P.R. of China, Germany, Hungary, Japan, Republic of Korea, Russian Federation, Switzerland, United Kingdom, and the United States of America. The IRPhEP Handbook is available to authorised requesters from the

  13. 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)

  14. 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)

  15. Design characteristics of research zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Nikolic, D.; Antic, D.; Zavaljevski, N.; Popovic, D.

    1990-01-01

    LASTA is a flexible zero power reactor with uranium and plutonium fuel designed for research in the neutron physics and in the fast reactor physics. Safety considerations and experimental flexibility led to the choice of a fixed vertical assembly with two safety blocks as the main safety elements, so that safety devices would be operated by gravity. The neutron and reactor physics, the control and safety philosophy adopted in our design, are described in this paper. Developed computer programs are presented. (author)

  16. Neutronics characteristics of space power reactors

    International Nuclear Information System (INIS)

    Little, W.; Barner, J.

    1986-01-01

    The objective of the paper is to describe the neutronic characteristics of a range of possible space reactor designs, and indicate the relative advantages and disadvantages of the various designs. Fuel designs to be considered are cermets (i.e., ceramic particles embedded in a metal matrix) consisting of UO 2 or Nn ceramic particles in matrices of Nb, Mo, Ta, or W. These cermet fuels are compared to a UN pin-type design. UN was selected for the reference fuel material since it has a somewhat higher density than UO 2 (i.e., 14.32 versus 10.96 gm/cc), which allows a lower minimum critical mass for both ceramic and cermet designs

  17. Expected characteristics of future reactors for human beings

    International Nuclear Information System (INIS)

    Taketani, Kiyoaki

    1992-01-01

    Based on four reactor safety components (namely: a) God-given safety, b) Equipment safety, c) Quick-response safety, d) Containing safety), categorical assessment is made of various nuclear reactor concepts ranging from present existing reactors to future reactors based on innovative reactor design. In pursuit of nuclear reactor safety, ultimate characteristics of the ideal nuclear reactor are expected to coincide with those of an inherently safe reactor. A definition of 'inherently safe' has already been proposed by a committee in Japan. As a realistic and existable reactor, which is as close to the ideal reactor, a future reactor which is almost the same as a global reactor, is proposed. This global reactor must be constructable anywhere on earth and must permit easy operation and maintenance by anyone. It is also discussed to identify what behavior is expected of the global reactor under various conditions. At the same time, this future reactor which includes the global reactor, should solve a) the nuclear fuel resource issue, b) efficient utilization of fission energy and c) environmental issues as the greenhouse effect. (author). 7 refs., 2 figs

  18. Introduction to nuclear power reactors and their health physics systems

    International Nuclear Information System (INIS)

    Brtis, J.S.

    1982-01-01

    This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

  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. Poly(L-lactic acid) microspheres containing neutron-activatable holmium-165: a study of the physical characteristics of microspheres before and after irradiation in a nuclear reactor.

    Science.gov (United States)

    Mumper, R J; Jay, M

    1992-01-01

    The solvent evaporation technique was employed to prepare poly(L-lactic acid) (PLA) microspheres with 165Ho acetylacetonate (Ho-AcAc). Particle size, percentage Ho-165, percent residual solvent, and retentive ability of the spheres were found to be strongly affected by preparatory conditions. Differential scanning calorimetry (DSC) thermograms suggested that the Ho-AcAc existed in the PLA matrix as a molecular dispersion. High neutron flux irradiations of the PLA spheres in a nuclear reactor produced Ho-166, a therapeutic radionuclide that emits high-energy negatrons (Emax = 1.84 MeV; half-life = 26.9 hr). The gamma radiation dose (53-75 Mrad) from the core of the reactor provided an overkill of all bioburdens in the PLA spheres. Gel permeation chromatography (GPC) analysis showed that these irradiations caused a reduction in PLA molecular weight. Infrared spectra, 13C NMR spectra, 1H NMR spectra, and DSC thermograms further confirmed the presence of lower molecular weight PLA but proved the overall maintenance of PLA structure.

  2. Research about reactor operator's personality characteristics and performance

    International Nuclear Information System (INIS)

    Wei Li; He Xuhong; Zhao Bingquan

    2003-01-01

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  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. Physical Therapy and Manual Physical Therapy: Differences in Patient Characteristics.

    NARCIS (Netherlands)

    van Ravensberg, C. D. Dorine; Oostendorp, Rob A B; van Berkel, Lonneke M.; Scholten-Peeters, Gwendolijne G M; Pool, Jan J.M.; Swinkels, Raymond A. H. M.; Huijbregts, Peter A.

    2005-01-01

    This study compared socio-demographic characteristics, health problem characteristics, and primary process data between database samples of patients referred to physical therapy (PT) versus a sample of patients referred to manual physical therapy (MPT) in the Netherlands. Statistical analysis

  5. Physical therapy and manual physical therapy: Differences in patient characteristics

    NARCIS (Netherlands)

    van Ravensberg, C. D. Dorine; Oostendorp, R.A.B.; van Berkel, Lonneke M.; Scholten-Peeters, G.G.M.; Pool, J.J.M.; Swinkels, Raymond A. H. M.; Huijbregts, Peter A.

    2005-01-01

    This study compared socio-demographic characteristics, health problem characteristics, and primary process data between database samples of patients referred to physical therapy (PT) versus a sample of patients referred to manual physical therapy (MPT) in the Netherlands. Statistical analysis

  6. Physical Therapy and Manual Physical Therapy: Differences in Patient Characteristics

    NARCIS (Netherlands)

    van Ravensberg, C. D. Dorine; Oostendorp, Rob A B; van Berkel, Lonneke M.; Scholten-Peeters, Gwendolijne G. M.; Pool, Jan J.M.; Swinkels, Raymond A. H. M.; Huijbregts, Peter A.

    2005-01-01

    This study compared socio-demographic characteristics, health problem characteristics, and primary process data between database samples of patients referred to physical therapy (PT) versus a sample of patients referred to manual physical therapy (MPT) in the Netherlands. Statistical analysis

  7. Physics of Plutonium Recycling in Thermal Reactors

    International Nuclear Information System (INIS)

    Kinchin, G.H.

    1967-01-01

    A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of 240 Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

  8. Physics of Plutonium Recycling in Thermal Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kinchin, G. H. [Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1967-09-15

    A substantial programme of experimental reactor physics work with plutonium fuels has been carried out in the UK; the purpose of this paper is to review the experimental and theoretical work, with emphasis on plutonium recycling in thermal reactors. Although the main incentive for some of the work may have been to study plutonium build-up in uranium-fuelled reactors, it is nevertheless relevant to plutonium recycling and no distinction is drawn between build-up and enrichment studies. A variety of techniques have been for determining reactivity, neutron spectrum and reaction rates in simple assemblies of plutonium-aluminium fuel with water, graphite and beryllia moderators. These experiments give confidence in the basic data and methods of calculation for near-homogeneous mixtures of plutonium and moderator. In the practical case of plutonium recycling it is necessary to confirm that satisfactory predictions can be made for heterogeneous lattices enriched with plutonium. In this field, experiments have been carried out with plutonium-uranium metal and oxide-cluster fuels in graphite-moderated lattices and in SGHW lattices, and the effects of {sup 240}Pu have been studied by perturbation measurements with single fuel elements. The exponential and critical experiments have used tonne quantities of fuel with plutonium contents ranging from 0.25 to 1.2% and the perturbation experiments have extended both the range of plutonium contents and the range of isotopic compositions of plutonium. In addition to reactivity and reactivity coefficients, such as the temperature coefficients, attention has been concentrated on relative reaction rate distributions which provide evidence for variations of neutron spectrum. .Theoretical comparisons, together with similar comparisons for non-uniform lattices, establish the validity of methods of calculation which have been used to study the feasibility of plutonium recycling in thermal reactors. (author)

  9. Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

    KAUST Repository

    Bin Hamzah, Muhamad Firdaus

    2017-05-01

    Ignition and combustion characteristics of n-butanol/air, 2-butanol.air and isobutanol/air mixtures at stoichiometric (ϕ = 1) and lean (ϕ = 0.5) conditions were investigated in a micro flow reactor with a controlled temperature profile from 323 K to 1313 K, under atmospheric pressure. Sole distinctive weak flame was observed for each mixture, with inlet fuel/air mixture velocity set low at 2 cm/s. One-dimensional computation with comprehensive chemistry and transport was conducted. At low mixture velocities, one-stage oxidation was confirmed from heat release rate profiles, which was broadly in agreement with the experimental results. The weak flame positions were congruent with literature describing reactivity of the butanol isomers. These weak flame responses were also found to mirror the trend in Anti-Knock Indexes of the butanol isomers. Flux and sensitivity analyses were performed to investigate the fuel oxidation pathways at low and high temperatures. Further computational investigations on oxidation of butanol isomers at higher pressure of 5 atm indicated two-stage oxidation through the heat release rate profiles. Low temperature chemistry is accentuated in the region near the first weak cool flame for oxidation under higher pressure, and its impact on key species – such as hydroxyl radical, hydrogen peroxide and carbon monoxide – were considered. Both experimental and computational findings demonstrate the advantage of employing the micro flow reactor in investigating oxidation processes in the temperature region of interest along the reactor channel. By varying physical conditions such as pressure, the micro flow reactor system is proven to be highly beneficial in elucidating oxidation behavior of butanol isomers in conditions in engines such as those that mirror HCCI operations.

  10. Lessons from feedback of safety operating experience for reactor physics

    International Nuclear Information System (INIS)

    Suchomel, J.; Rapavy, S.

    1999-01-01

    Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

  11. Applications in nuclear data and reactor physics

    International Nuclear Information System (INIS)

    Cullen, D.E.; Muranaka, R.; Schmidt, J.

    1986-01-01

    This book presents the papers given at a conference on reactor kinetics and nuclear data collections. Topics considered at the conference included nuclear data processing, PWR core design calculations, reactor neutron dosimetry, in-core fuel management, reactor safety analysis, transients, two-phase flow, fuel cycles of research reactors, slightly enriched uranium, highly enriched uranium, reactor start-up, computer codes, and the transport of spent fuel elements

  12. Advances in Reactor Physics, Mathematics and Computation. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

  13. 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)

  14. 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

  15. Schoolyard Characteristics, Physical Activity, and Sedentary Behavior

    DEFF Research Database (Denmark)

    Van Kann, Dave H H; de Vries, Sanne I; Schipperijn, Jasper

    2016-01-01

    BACKGROUND: Physical activity (PA) is decreasing among children, while sedentary behavior (SB) is increasing. Schoolyards seem suitable settings to influence children's PA behavior. This study investigated the associations between schoolyard characteristics and moderate-to-vigorous physical activ...

  16. Reactor physics aspects of burning actinides in a nuclear reactor

    International Nuclear Information System (INIS)

    Hage, W.; Schmidt, E.

    1978-01-01

    A short review of the different recycling strategies of actinides other than fuel treated in the literature, is given along with nuclear data requirements for actinide build-up and transmutation studies. The effects of recycling actinides in a nuclear reactor on the flux distribution, the infinite neutron multiplication factor, the reactivity control system, the reactivity coefficients and the delayed neutron fraction are discussed considering a notional LWR or LMFBR as an Actinide Trasmutaton Reactor. Some operational problems of Actinide Transmutation reactors are mentioned, which are caused by the α-decay heat and the neutron sources of Actinide Target Elements

  17. 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)

  18. 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.)

  19. First physical start-up for the first pulsed reactor in China

    International Nuclear Information System (INIS)

    Huang Wenlou; Tan Rilin; Xie Yuqi; Chai Songshan; Li Yingfa; He Qianming; Zhou Bin

    1993-01-01

    The characteristics and the test results of initial loading fuel and first physical start-up for the first pulsed reactor in China (PRC-1) are described. Safe measure to ensure safety of first physical start-up are also described. The experiments show that performances of PRC-1 are in accord with design requirements

  20. 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)

  1. A Study on the Kinetic Characteristics of Transmutation Process Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae seon; Huh, Chang Wook; Kim, Doh Hyung [Seoul National University, Seoul (Korea, Republic of)

    1997-07-01

    The purpose of this study is to examine the transient heat transfer characteristics of liquid mental as the coolant used in accelerator-driven transmutation process reactor which is related the disposal of high-level radioactive nuclide. At current stage, the accelerator-driven transmutation process is investigated as the most appropriate method among many transmutation process methods. In this study, previous research works are investigated especially about the thermal hydraulics and kinetic behavior of coolant material including heat transfer of coolant in transmutation process reactor. A study on the heat transfer characteristics of liquid metal is performed based on the thermal hydraulic kinetic characteristics of liquid metal reactor which uses liquid metal coolant. Based on this study, the most appropriate material for the coolant of transmutation reactor will be recommended. 53 refs., 15 tabs., 33 figs. (author)

  2. 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

  3. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1994-01-01

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  4. Research on acceleration method of reactor physics based on FPGA platforms

    International Nuclear Information System (INIS)

    Li, C.; Yu, G.; Wang, K.

    2013-01-01

    The physical designs of the new concept reactors which have complex structure, various materials and neutronic energy spectrum, have greatly improved the requirements to the calculation methods and the corresponding computing hardware. Along with the widely used parallel algorithm, heterogeneous platforms architecture has been introduced into numerical computations in reactor physics. Because of the natural parallel characteristics, the CPU-FPGA architecture is often used to accelerate numerical computation. This paper studies the application and features of this kind of heterogeneous platforms used in numerical calculation of reactor physics through practical examples. After the designed neutron diffusion module based on CPU-FPGA architecture achieves a 11.2 speed up factor, it is proved to be feasible to apply this kind of heterogeneous platform into reactor physics. (authors)

  5. Advances in Reactor Physics, Mathematics and Computation. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

  6. An overview of reactor physics standards: Past, present and future

    International Nuclear Information System (INIS)

    Cokinos, D.M.

    1992-07-01

    This report discusses for determining key static reactor physics parameters which have been developed by groups of experts (working groups) under the aegis of ANS-19, the ANS Reactor Physics Standards Committee. Following a series of sequential reviews, augmented by feedback from potential users, a proposed standard is brought into final form by the working group before it is adopted as a formal standard by the American National Standards Institute (ANSI); Reactor Physics standards are intended to provide guidance in the performance and qualification of complex sequences of reactor calculations and/or measurements and are regularly reviewed for possible updates and/or revisions. The reactor physics standards developed to date are listed and standards now being developed by the respective working groups are also provided

  7. Zero energy reactor RB technical characteristics and experimental possibilities

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, S; Takac, S; Raisic, N; Lolic, B; Markovic, H [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1963-04-15

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility.

  8. Zero energy reactor RB technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Jovanovic, S.; Takac, S.; Raisic, N.; Lolic, B.; Markovic, H.

    1963-04-01

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility

  9. Nuclear Data Processing for Reactor Physics Calculation

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Pandiangan, Tumpal

    2003-01-01

    Nuclear data processing for reactor physics calculation has been done. Raw nuclear data cross-sections on file ENDF should be prepared and processed before it used in neutronic calculation. The processing code system such as NJOY-PC code has been used from linearization of nuclear cross-sections data and background contribution of resonance parameter (MF2) using RECONR module (0K) with energy range from 10 -5 to 10 7 eV. Afterward, the neutron cross-sections data should be processed and broadened to desire temperature (i.e. 293K) by using BROADR module. The Grouper and Therma modules will be applied for multi-groups calculation which suitable for WIMS/D4 (69 groups) and thermalization of nuclear constants. The final stage of processing nuclear cross-sections is updating WIMS/D4 library. The WIMSR module in NJOY-PC and WILLIE code will be applied in this stage. The evaluated nuclear data file, especially for 1 H 1 isotope, was taken from JENDL-3.2 and ENDF/B-VI for preliminary study. The results of nuclear data processing 1 H 1 shows that the old-WIMS (WIMS-lama) library have much discrepancies comparing with JENDL-3.2 or ENDF/B-VI files, especially in energy around 5 keV

  10. Reactor physics special problem in 11. ENFIR

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    1997-01-01

    In this report, the computation method and the results of the work performed of the special topic on reactor physics proposed for the 11. ENFIR is presented. MCNP 4.2 has been adopted as the only code to perform the calculations. The full core of the IPEN-MB-1 critical unit has been modelled without important approximations. The specifications given by the Organizer Commission of the Special Topic were followed. The nuclear libraries adopted were those included on the MCNPDAT package, mainly from ENDF/B-V, except indium data, not included in this package. For indium, data obtained from LANL, based on ENDF/B-VI were used. The results are: critical position of the control banks assuming simultaneous movement: percent of extraction: (49±1)% ; excess of reactivity of the core: ρ =( 3590 ±50)pcm ; total reactivity of the one control rod bank: ρ= (4000±50) pcm. The reactivity curve of the control rods is included also. (author)

  11. Burnup characteristics of binary breeder reactors

    International Nuclear Information System (INIS)

    Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.

    1983-01-01

    Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt

  12. Reactor physics and thermodynamics of a gaseous core fission reactor

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Van Dam, H.; Stekelenburg, A.J.C.; Hoogenboom, J.E.; Boersma-Klein, W.; Kistemaker, J.

    1990-01-01

    Neutron kinetics and thermodynamics of a Gaseous Core Fission Reactor with magnetical pumping are shown to have many unconventional aspects. Attention is focussed on the properties of the fuel gas, the stationary temperature distribution, the non-linear neutron kinetics and the energy balance in thermodynamical cycles

  13. 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

  14. Nuclear characteristics of D-D fusion reactor blankets

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao

    1978-01-01

    Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

  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. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  17. Twenty years of health physics research reactor operation

    International Nuclear Information System (INIS)

    Sims, C.S.; Gilley, L.W.

    1983-01-01

    The Health Physics Research Reactor at the Oak Ridge National Laboratory has been in regular use for more than two decades. Safe operation of this fast reactor over this extended period indicates that (1) fundamental design, (2) operational procedures, (3) operator training and performance, (4) maintenance activites, and (5) management have all been eminently satisfactory. The reactor and its uses are described, the operational history and significant events are reviewed, and operational improvements and maintenance are discussed

  18. Economic characteristics of a smaller, simpler reactor

    International Nuclear Information System (INIS)

    LaBar, M.; Bowers, H.

    1988-01-01

    Reduced load growth and heightened concern with economic risk has led to an expressed utility preference for smaller capacity additions. The Modular High Temperature Reactor (MHTGR) plant has been developed as a small, simple plant that has limited financial risk and is economically competitive with comparatively sized coal plants. Competitive economics is achieved by the simplifications made possible in a small MHTGR, reduction in the quantity of nuclear grade construction and design standardization and certification. Assessments show the MHTGR plant to have an economic advantage over coal plants for plant sizes from 270 MWe to 1080 MWe. Financial risk is limited by small unit sizes and short lead times that allow incremental deployment. Evaluations show the MHTGR incremental deployment capability to reduce negative cash flows by almost a factor of 2 relative to that required by a single large nuclear plant

  19. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    International Nuclear Information System (INIS)

    Moiseyev, A.V.

    2008-01-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  20. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  1. Reactor physics activities in France. October 1983 - September 1984

    International Nuclear Information System (INIS)

    Golinelli, C.; Salvatores, M.

    1984-10-01

    The major activities of the Fast Reactor Physics Program during the period October 1983 - September 1984 are reviewed: experimental and theoretical studies, computer codes. The LWR program brought improvements in the field of the Advanced Reactors and of the plutonium re-use on French PWRs. Are reviewed experimental studies and facilities, theoretical studies (transport theory, radioactive decay library)

  2. 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

  3. Parameter analysis calculation on characteristics of portable FAST reactor

    International Nuclear Information System (INIS)

    Otsubo, Akira; Kowata, Yasuki

    1998-06-01

    In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

  4. Reactor Physics Behind the Chernobyl Accident

    International Nuclear Information System (INIS)

    Reisch, F.

    1999-01-01

    There are some fourteen Chernobyl type of power reactors (1000 MWe) in operation at five different sites in Eastern Europe. In Russia; in St. Petersburg (4). in Smolensk (3). and in Kursk (4) in the Ukraine in Chernobyl (l) and in Lithuania in Ignalina (2). The oldest one is west of St. Petersburg and the most powerful one is in Ignalina. The reactors at St. Petersburg and in Lithuania are near to the Baltic sea. An intricate reactor construction was the most important cause of the accident. There were other reasons too: human error. politics and economics

  5. Neutron transport. Physics and calculation of nuclear reactors with applications to pressurized water reactors and fast neutron reactors. 2 ed.

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

    This book presents the main physical bases of neutron theory and nuclear reactor calculation. 1) Interactions of neutrons with matter and basic principles of neutron transport; 2) Neutron transport in homogeneous medium and the neutron field: kinetic behaviour, slowing-down, resonance absorption, diffusion equation, processing methods; 3) Theory of a reactor constituted with homogeneous zones: critical condition, kinetics, separation of variables, calculation and neutron balance of the fundamental mode, one-group and multigroup theories; 4) Study of heterogeneous cell lattices: fast fission factor, resonance absorption, thermal output factor, diffusion coefficient, computer codes; 5) Operation and control of reactors: perturbation theory, reactivity, fuel properties evolution, poisoning by fission products, calculation of a reactor and fuel management; 6) Study of some types of reactors: PWR and fast breeder reactors, the main reactor types of the present French program [fr

  6. Compilation of reactor physics data of the year 1984, AVR reactor

    International Nuclear Information System (INIS)

    Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

    1985-12-01

    The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1984 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

  7. Compilation of reactor physics data of the year 1983, AVR reactor

    International Nuclear Information System (INIS)

    Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

    1985-06-01

    The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1983 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

  8. Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

    2003-12-01

    Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

  9. Fissile fuel doubling time characteristics for reactor lifetime fuel logistics

    International Nuclear Information System (INIS)

    Heindler, M.; Harms, A.A.

    1978-01-01

    The establishment of nuclear fuel requirements and their efficient utilization requires a detailed knowledge of some aspects of fuel dynamics and processing during the reactor lifetime. It is shown here that the use of the fuel stockpile inventory concept can serve effectively for this fuel management purpose. The temporal variation of the fissile fuel doubling time as well as nonequilibrium core conditions are among the characteristics which thus become more evident. These characteristics - rather than a single figure-of-merit - clearly provide an improved description of the expansion capacity and/or fuel requirements of a nuclear reactor energy system

  10. 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)

  11. Occupational health physics at a fusion reactor

    International Nuclear Information System (INIS)

    Shank, K.E.; Easterly, C.E.; Shoup, R.L.

    1975-01-01

    Future generation of electrical power using controlled thermonuclear reactors will involve both traditional and new concerns for health protection. A review of the problems associated with exposures to tritium and magnetic fields is presented with emphasis on the occupational worker. The radiological aspects of tritium, inventories and loss rates of tritium for fusion reactors, and protection of the occupational worker are discussed. Magnetic fields in which workers may be exposed routinely and possible biological effects are also discussed

  12. Current status of fast reactor physics

    International Nuclear Information System (INIS)

    Hummel, H.H.

    1979-01-01

    The subject of calculation of reactivity coefficients for fast reactors is developed, starting with a discussion of the status of relevant nuclear data and proceeding to the subjects of group cross section generation and of methods of obtaining reactivity coefficients from group cross sections. Reactivity coefficients measured in critical experiments are compared with calculated values. Dependence of reactivity coefficients on reactor design is discussed. Finally, results of the recent international comparison of calculated reactivity coefficients are presented

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

    International Nuclear Information System (INIS)

    2011-01-01

    Full text : The international conference on physics and technology of reactors is organized by the Moroccan Association for Nuclear enggineering and Reactor Technology (GMTR) with the collaboration of the Centre for Energy and Nuclear Sciences and Techniques (CNESTEN) and under the auspices of the ministry of Energy, Mining, Water and Environment. The programme of the PHYTRA2 conference covers a wide variety of topics. The conference is organised in one plenary session, eight oral technical sessions and one poster session. The oral and poster technical sessions covers the usual topics of nuclear engineering including one session on research reactors utilisation and computational methods for research reactors

  14. Experimental Equipment for Physics Studies in the Aagesta Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bernander, G; Blomberg, P E; Dubois, P O

    1967-03-15

    Comprehensive physics measurements were carried out in connection with the start up of the Agesta reactor. For this purpose special experimental equipment was constructed and installed in the reactor. Parts of this were indispensable and/or time-saving for the reactivity control during the core build-up period and during the first criticality studies. This report gives mainly a detailed description of the experimental equipment used, but also the relevant physics background and the experience gained during the performance.

  15. Characteristics of outage radiation fields around various reactor components

    International Nuclear Information System (INIS)

    Verzilov, Y.; Husain, A.; Corbin, G.

    2008-01-01

    Full text: Activity monitoring surveys, consisting of gamma spectroscopy and dose rate measurements, of various CANDU station components such as the reactor face, feeder cabinet, steam generators and moderator heat exchangers are often performed during shutdown in order to trend the transport of activity around the primary heat transport and moderator systems. Recently, the increased dose expenditure for work such as feeder inspection and replacement in the reactor vault has also spurred interest in improved characterization of the reactor face fields to facilitate better ALARA decision making and hence a reduction in future dose expenditures. At present, planning for reactor face work is hampered by insufficient understanding of the relative contribution of the various components to the overall dose. In addition to the increased dose expenditure for work at the reactor face, maintenance work associated with horizontal flux detectors and liquid injection systems has also resulted in elevated dose expenditures. For instance at Darlington, radiation fields in the vicinity of horizontal flux detectors (HFD) and Liquid Injection Shutdown System (LISS) nozzle bellows are trending upwards with present contact fields being in the range 16-70 rem/h and working distance fields being in the range 100-500 mrem/h. This paper presents findings based on work currently being funded by the CANDU Owners Group. Measurements were performed at Ontario Power Generation's Pickering and Darlington nuclear stations. Specifically, the following are addressed: Characteristics of Reactor Vault Fields; Characteristics of Steam Generator Fields; Characteristics of Moderator Heat Exchanger Fields. Measurements in the reactor vault were performed at the reactor face, along the length of end fittings, along the length of feeders, at the bleed condenser and at the HFD and LISS nozzle bellows. Steam generator fields were characterized at various elevations above the tube sheet, with and without the

  16. Safety characteristics of the integral fast reactor concept

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Cahalan, J.E.; Sevy, R.H.; Wright, A.E.

    1985-01-01

    The Integral Fast Reactor (IFR) concept is an innovative approach to liquid metal reactor design which is being studied by Argonne National Laboratory. Two of the key features of the IFR design are a metal fuel core design, based on the fuel technology developed at EBR-II, and an integral fuel cycle with a colocated fuel cycle facility based on the compact and simplified process steps made possible by the use of metal fuel. The paper presents the safety characteristics of the IFR concept which derive from the use of metal fuel. Liquid metal reactors, because of the low pressure coolant operating far below its boiling point, the natural circulation capability, and high system heat capacities, possess a high degree of inherent safety. The use of metallic fuel allows the reactor designer to further enhance the system capability for passive accommodation of postulated accidents

  17. 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

  18. Physical characteristics of satellite surfaces

    International Nuclear Information System (INIS)

    Veverka, J.; Thomas, P.; Johnson, T.V.; Matson, D.; Housen, K.

    1986-01-01

    Both exogenic and endogenic effects have been proposed to explain the major observed characteristics of satellite surfaces. The current view is that the basic properties of most surfaces result from the intrinsic composition of a body and its geologic history. Exogenic effects have, however, played a role in modifying the appearance of nearly all surfaces. The most important exogenic effect is impact cratering, one manifestation of which is the production of micrometeoroid gardened regoliths on airless bodies. On large, silicate bodies the micrometeoroid bombardment can produce an optically mature, dark agglutinate-rich soil; the nature of regoliths on predominantly icy satellites remains uncertain. Direct accumulation of infalling material does not appear to play a major role in modifying most surfaces. Solar wind radiation effects have not altered greatly the optical properties of solar system objects; magnetospheric charged particles may have modified the optical properties of some outer planet satellites (e.g., sulfur ion bombardment in the case of some of the satellites of Jupiter). Other effects, such as aeolian and liquid/solid chemical weathering, may be important on satellites with atmospheres like Titan and Triton

  19. Photodegradation of polymers physical characteristics and applications

    CERN Document Server

    Rabek, Jan F

    1996-01-01

    In this book on physical characteristics and practical aspects of polymer photodegradation Rabek emphasizes the experimental work on the subject. The most important feature of the book is the physical interpretation of polymer degradation, e.g. mechanism of UV/light absorption, formation of excited states, energy transfer mechanism, kinetics, dependence on physical properties of macromolecules and polymer matrices, formation of mechanical defects, practics during environmental ageing. He includes also some aspects of polymer photodegradation in environmental and space condition.

  20. Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

    Energy Technology Data Exchange (ETDEWEB)

    Gutmann, H

    1972-06-05

    The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

  1. 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.

  2. 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.

  3. Characteristics of Butanol Isomers Oxidation in a Micro Flow Reactor

    KAUST Repository

    Bin Hamzah, Muhamad Firdaus

    2017-01-01

    Ignition and combustion characteristics of n-butanol/air, 2-butanol.air and isobutanol/air mixtures at stoichiometric (ϕ = 1) and lean (ϕ = 0.5) conditions were investigated in a micro flow reactor with a controlled temperature profile from 323 K

  4. Kinetic characteristics of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    An, Tran Khac; Dien, Nguyen Nhi; Hien, Pham Duy [Nuclear Research Inst., Da Lat (Viet Nam); and others

    1994-10-01

    Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated. Experimental parameters measured consist of: temperature coefficient of reactivity for water moderator, xenon poisoning, contribution of delayed photoneutrons induced by Be({gamma}, n) reactions and positive reactivity insertion behavior. (author). 6 refs. 4 figs.

  5. Kinetic characteristics of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Tran Khac An; Nguyen Nhi Dien; Pham Duy Hien

    1994-01-01

    Kinetic characteristics of the reconstructed nuclear reactor in Dalat is investigated. Experimental parameters measured consist of: temperature coefficient of reactivity for water moderator, xenon poisoning, contribution of delayed photoneutrons induced by Be(γ, n) reactions and positive reactivity insertion behavior. (author). 6 refs. 4 figs

  6. Physics experiments with the operating reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cullington, G R; King, D C

    1973-09-27

    Experimental techniques have been developed and used on Dragon to give consistent information on excess reactivity and shut down margin. The reactivity measurements have been correlated with the theoretical calculations and have led to improvements in the calculations. The methods used and the results obtained are accepted by the Safety Committee as sufficient evidence for compliance with the fuel loading safety rules. Although the reactor was not designed as an experimental facility, flux and dose measurements experiments have been successfully carried out. Mass flow and negative reactivity transient measurements have been carried out. These are valuable for demonstration of the flexibility of the reactor system and for giving confidence in theoretical calculations.

  7. Physics experiment on the Dragon reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, C.

    1974-10-15

    The paper describes a set of DRAGON experiments planned to measure burn-up effects in DRAGON irradiated fuel. Irradiated fuel elements from DRAGON are to be subjected to reactivity measurements in the HECTOR experimental reactor to infer the residual U235 content followed by isotopic analyses at CEA laboratories in 1975. Fast neutron damage to DRAGON graphite is compared to fast neutron dose measurements using Ni58 (n,p) Co58 activation wires in both DRAGON and the DIDO MTR. Gamma scanning of irradiated fuel elements are used to compare axial power profiles to those derived from two-dimensional and three-dimensional calculations of the DRAGON reactor.

  8. Physical characteristics comparison of virtual wedge device with physical wedge

    International Nuclear Information System (INIS)

    Cho, Jung Keun; Choi, Kye Sook; Lim, Cheong Hwan; Kim, Jeong Koo; Jung, Hong Ryang; Lee, Jung Ok; Lee, Man Goo

    2001-01-01

    We compared the characteristics of Siemens virtual wedge device with physical wedges for clinical application. We investigated the characteristics of virtual and physical wedges for various wedge angles (15, 30, 45, and 60) using 6- and 15- MV photon beams. Wedge factors were measured in water using an ion chamber for various field sizes and depths. In case of virtual wedge device, as upper jaw moves during irradiation, wedge angles were estimated by accumulated doses. These measurements were performed at off-axis points perpendicular to the beam central axis in water for a 15 cm x 20 cm radiation field size at the depth of 10 cm. Surface does without and with virtual or physical wedges were measured using a parallel plate ion chamber at surface. Field size was 15 cm * 20 cm and a polystyrene phantom was used. For various field sizes, virtual and physical wedge factors were changed by maximum 2.1% and 3.9%, respectively. For various depths, virtual and physical wedge factors were changed by maximum 1.9% and 2.9%, respectively. No major difference was found between the virtual and physical wedge angles and the difference was within 0.5. Surface dose with physical wedge was reduced by maximum 20% (x-ray beam : 6 MV, wedge angle : 45, SSD : 80 cm) relative to one with virtual wedge or without wedge. Comparison of the characteristics of Siemens virtual wedge device with physical wedges was performed. Depth dependence of virtual wedge factor was smaller than that of physical wedge factor. Virtual and physical wedge factors were nearly independent of field sizes. The accuracy of virtual and physical wedge angles was excellent. Surface dose was found to be reduced using a physical wedge

  9. Physical characteristics comparison of virtual wedge device with physical wedge

    International Nuclear Information System (INIS)

    Choi, Dong Rak; Shin, Kyung Hwan; Lee, Kyu Chan; Kim, Dae Yong; Ahn, Yong Chan; Lim, Do Hoon; Kim, Moon Kyung; Huh, Seung Jae

    1999-01-01

    We have compared the characteristics of Siemens virtual wedge device with physical wedges for clinical application. We investigated the characteristics of virtual and physical wedges for various wedge angles (15, 30, 45, and 60 ) using 6- and 15MV photon beams. Wedge factors were measured in water using an ion chamber for various field sizes and depths. In case of virtual wedge device, as upper jaw moves during irradiation, wedge angles were estimated by accumulated doses. These measurements were performed at off-axis points perpendicular to the beam central axis in water for a 15 cm x 20 cm radiation field size at the depth of 10 cm. Surface doses without and with virtual or physical wedges were measured using a parallel plate ion chamber at surface. Field size was 15 cm x 20 cm and a polystyrene phantom was used. For various field sizes, virtual and physical wedge factors were changed by maximum 2.1% and 3.9%, respectively. For various depths, virtual and physical wedge factors were changed by maximum 1.9% and 2.9%, respectively. No major difference was found between the virtual and physical wedge angles and the difference was within 0.5 .deg. . Surface dose with physical wedge was reduced by maximum 20% (x-ray beam: 6 MV, wedge angle: 45 .deg. , SSD: 80cm) relative to one with virtual wedge or without wedge. Comparison of the characteristics of Siemens virtual wedge device with physical wedges was performed. Depth dependence of virtual wedge factor was smaller than that of physical wedge factor. Virtual and physical wedge factors were nearly independent of field sizes. The accuracy of virtual and physical wedge angles was excellent. Surface dose was found to be reduced using physical wedge

  10. Safety characteristics of the US advanced liquid metal reactor core

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

    1991-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

  11. Physical characteristics related to bra fit.

    Science.gov (United States)

    Chen, Chin-Man; LaBat, Karen; Bye, Elizabeth

    2010-04-01

    Producing well-fitting garments has been a challenge for retailers and manufacturers since mass production began. Poorly fitted bras can cause discomfort or pain and result in lost sales for retailers. Because body contours are important factors affecting bra fit, this study analyses the relationship of physical characteristics to bra-fit problems. This study has used 3-D body-scanning technology to extract upper body angles from a sample of 103 college women; these data were used to categorise physical characteristics into shoulder slope, bust prominence, back curvature and acromion placement. Relationships between these physical categories and bra-fit problems were then analysed. Results show that significant main effects and two-way interactions of the physical categories exist in the fit problems of poor bra support and bra-motion restriction. The findings are valuable in helping the apparel industry create better-fitting bras. STATEMENT OF RELEVANCE: Poorly fitted bras can cause discomfort or pain and result in lost sales for retailers. The findings regarding body-shape classification provide researchers with a statistics method to quantify physical characteristics and the findings regarding the relationship analysis between physical characteristics and bra fit offer bra companies valuable information about bra-fit perceptions attributable to women with figure variations.

  12. Brief history of the reactor physics activities at ICN Pitesti

    International Nuclear Information System (INIS)

    Dumitrache, I.

    2004-01-01

    The Institute was established 33 years ago, in April 1971. Several specialists from the Institute for Atomic Physics - Bucharest came at the new research entity and the reactor physics activities had a successful start. One can identify three distinct periods: 1971-1980, the Bucharest years, 1980-1996, solving critical problems years and 1977-present (2004), technical support years. The first period is usually seen as a training one. This is only partially true. Most of the physicists came from University in 1971 and 1972 years. A significant number of them were trained abroad, in France, Germany, Italy, USA, Canada etc., usually under IAEA Vienna fellowships. The work was really pleasant and the progress was exciting. Unfortunately, the main task (to design a thermal reactor and a fast reactor, both for research activities) was, probably, much too difficult from the technical point of view and, in addition, required an unrealistic economic effort. In the Fall of the 1976 year, most of the reactor physicists were removed from Bucharest to Pitesti. One year later, all the remaining specialists were concentrated in Pitesti. The dual core TRIGA reactors were commissioned in the last months of the 1979 year. The CYBER 720 mainframe computer was available in December 1980. Between 1980 and 1992 years, practically all the Romanian activities related to reactor physics were performed in Pitesti, Mioveni compound. The details related to critical problems will be presented in the paper. We mention here four of the problems that have a significant impact even today, namely: -Final dimensioning of the adjuster rods for the Cernavoda NPP, Unit 2. The rods were manufactured in USA and Canada, using the AECL design and the final dimensions have been specified by ICN Pitesti; -Use of the LEU fuel in TRIGA-SSR Reactor, instead of the original HEU fuel; -Design of the irradiation experiments in TRIGA cores, in order to provide the required conditions during the test, according to

  13. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  14. 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.)

  15. Research and development of a super fast reactor (12). Considerations for the reactor characteristics

    International Nuclear Information System (INIS)

    Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

    2008-01-01

    A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

  16. 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

  17. Physical properties of organic nuclear reactor coolants

    Energy Technology Data Exchange (ETDEWEB)

    Elberg, S.; Friz, G.

    1963-03-15

    Diphenyl and terphenyls with different high-boiler content were studied up to temperatures of 450 deg C. Data from high boiler reactors show viscosity (strong influence), thermal conductivity (medium influence), density and specific heat (small influence). The vapor pressure is rn the most affected property (important influence of low boilers). Also viscosity shows an effect. Some data for pure highboilers are also presented. New results were obtained with direct measurements of the latent heat ot vaporization. (P.C.H.)

  18. Operating manual for the Health Physics Research Reactor

    International Nuclear Information System (INIS)

    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

  19. Characteristics of the General Physics student population.

    Science.gov (United States)

    Hunt, Gary L.

    2006-12-01

    Are pre-medical students different than the other students in a General physics class? They often appear to be different, based on how often they seek help from the instructor or how nervous they are about 2 points on a lab report. But are these students different in a measurable characteristic? The purpose of this study is to better understand the characteristics of the students in the introductory physics classes. This is the first step toward improving the instruction. By better understanding the students the classroom, the organization and pedagogy can be adjusted to optimize student learning. The characteristics to be investigated during this study are: · student epistemological structure, · student attitudes, · science course preparation prior to this course, · study techniques used, · physics concepts gained during the class · performance in the class. The data will be analyzed to investigate differences between groups. The groups investigated will be major, gender, and traditional/nontraditional students.

  20. 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)

  1. The safety characteristics of the HTR 500 reactor plant

    International Nuclear Information System (INIS)

    Wachholz, W.

    1987-01-01

    The HTR is a reactor having a passive safety. It is equipped with the usual active engineered safety systems in simplified form. Due to its inherent safety characteristics and the burst-safe prestressed concrete reactor vessel activity containment is ensured even without the effect of active safety systems. Even in the event of extremely hypothetical accidents the effect on the environment is low enough so that evacuation or relocation of the population is not required. Therefore large-scale damage of agricultural land and industrially used areas is safely ruled out. Thus the site selection for this type of reactor is not restricted i.e. an HTR can be constructed near industrial and urban center. (author)

  2. Primary circuit and reactor core T-H characteristics determination of WWER 440 reactors

    International Nuclear Information System (INIS)

    Hermansky, J.; Petenyi, V.; Zavodsky, M.

    2010-01-01

    The WWER-440 nuclear fuel vendor permanently improves the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. During unit refuelling there also could be made some other changes in hydraulic parameters of primary circuit (change of impeller wheels, hydraulic resistance coefficient changes of internal parts of primary circuit, etc.). Therefore it is necessary to determine real coolant flow rate through the reactor during units start-up after their refuelling, and also to have the skilled methodology and computing code for analyzing factors, which affecting the inaccuracy of coolant flow redistribution determination through reactor on flows through separate parts of reactor core in any case of parallel operation of different assembly types. Computing code TH-VCR and CORFLO are used for reactor core characteristics determination for one type of fuel and control assemblies and also in case of parallel operation of different assembly types. The code TH-VCR is able to calculate coolant flow rate for different combinations of three different fuel assembly channel types and three different control assembly channel types. The CORFLO code deals the area of the reactor core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is taken into account. Computing code CORFLO is verified and validated at this time. Paper presents some results from measurements of coolant flow rate through reactors during start-up after unit refuelling and short description of computing code TH-VCR and CORFLO with some calculated results. (Authors)

  3. 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

  4. Development of the evaluation methods in reactor safety analyses and core characteristics

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to support the safety reviews by NRA on reactor safety design including the phenomena with multiple failures, the computer codes are developed and the safety evaluations with analyses are performed in the areas of thermal hydraulics and core characteristics evaluation. In the code preparation of safety analyses, the TRACE and RELAP5 code were prepared to conduct the safety analyses of LOCA and beyond design basis accidents with multiple failures. In the core physics code preparation, the functions of sensitivity and uncertainty analysis were incorporated in the lattice physics code CASMO-4. The verification of improved CASMO-4 /SIMULATE-3 was continued by using core physics data. (author)

  5. Physics methods for calculating light water reactor increased performances

    International Nuclear Information System (INIS)

    Vandenberg, C.; Charlier, A.

    1988-01-01

    The intensive use of light water reactors (LWRs) has induced modification of their characteristics and performances in order to improve fissile material utilization and to increase their availability and flexibility under operation. From the conceptual point of view, adequate methods must be used to calculate core characteristics, taking into account present design requirements, e.g., use of burnable poison, plutonium recycling, etc. From the operational point of view, nuclear plants that have been producing a large percentage of electricity in some countries must adapt their planning to the need of the electrical network and operate on a load-follow basis. Consequently, plant behavior must be predicted and accurately followed in order to improve the plant's capability within safety limits. The Belgonucleaire code system has been developed and extensively validated. It is an accurate, flexible, easily usable, fast-running tool for solving the problems related to LWR technology development. The methods and validation of the two computer codes LWR-WIMS and MICROLUX, which are the main components of the physics calculation system, are explained

  6. An optimization method for parameters in reactor nuclear physics

    International Nuclear Information System (INIS)

    Jachic, J.

    1982-01-01

    An optimization method for two basic problems of Reactor Physics was developed. The first is the optimization of a plutonium critical mass and the bruding ratio for fast reactors in function of the radial enrichment distribution of the fuel used as control parameter. The second is the maximization of the generation and the plutonium burnup by an optimization of power temporal distribution. (E.G.) [pt

  7. 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)

  8. 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

  9. Characteristics of fast reactor core designs and closed fuel cycle

    International Nuclear Information System (INIS)

    Poplavsky, V.M.; Eliseev, V.A.; Matveev, V.I.; Khomyakov, Y.S.; Tsyboulya, A.M.; Tsykunov, A.G.; Chebeskov, A.N.

    2007-01-01

    On the basis of the results of recent studies, preliminary basic requirements related to characteristics of fast reactor core and nuclear fuel cycle were elaborated. Decreasing reactivity margin due to approaching breeding ratio to 1, requirements to support non-proliferation of nuclear weapons, and requirements to decrease amount of radioactive waste are under consideration. Several designs of the BN-800 reactor core have been studied. In the case of MOX fuel it is possible to reach a breeding ratio about 1 due to the use of larger size of fuel elements with higher fuel density. Keeping low axial fertile blanket that would be reprocessed altogether with the core, it is possible to set up closed fuel cycle with the use of own produced plutonium only. Conceptual core designs of advanced commercial reactor BN-1800 with MOX and nitride fuel are also under consideration. It has been shown that it is expedient to use single enrichment fuel core design in this reactor in order to reach sufficient flattening and stability of power rating in the core. The main feature of fast reactor fuel cycle is a possibility to utilize plutonium and minor actinides which are the main contributors to the long-living radiotoxicity in irradiated nuclear fuel. The results of comparative analytical studies on the risk of plutonium proliferation in case of open and closed fuel cycle of nuclear power are also presented in the paper. (authors)

  10. DUPIC fuel performance from reactor physics viewpoint

    International Nuclear Information System (INIS)

    Choi, H.; Rhee, B.W.; Park, H.

    1995-01-01

    A preliminary study was performed for the evaluation of Stress Corrosion Cracking (SCC) parameters of nominal DUPIC fuel in CANDU reactor. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increase of the 43-element DUPIC fuel in the equilibrium core are below the SCC thresholds of CANDU natural uranium fuel. For 4-bundle shift refueling scheme, the envelope of element ramped power and power increase upon refueling are 8% and 44% higher than those of 2-bundle shift refueling scheme on the average, respectively, and both schemes are not expected to cause SCC failures. (author)

  11. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  12. CHARACTERISTICS OF CORN STALK HEMICELLULOSE PYROLYSIS IN A TUBULAR REACTOR

    OpenAIRE

    Gao-Jin Lv; Shu-Bin Wu; Rui Lou

    2010-01-01

    Pyrolysis characteristics of corn stalk hemicellulose were investigated in a tubular reactor at different temperatures, with focus mainly on the releasing profiles and forming behaviors of pyrolysis products (gas, char, and tar). The products obtained were further identified using various approaches (including GC, SEM, and GC-MS) to understand the influence of temperature on product properties and compositions. It was found that the devolatilization of hemicellulose mainly occurred at low tem...

  13. XII seminar on problems of reactor physics

    International Nuclear Information System (INIS)

    Kryuchkov, Eh.F.; Naumov, V.I.

    2003-01-01

    Results of the XII seminar Physical problems of effective and safety use of nuclear materials taking place on the basis of MEPI (September, 2002) are discussed. Reports on the directions: physical problems of advanced nuclear-energetic technologies; account, control and nuclear material management; effective and safety use of nuclear materials at NPP; programming and software for the analysis of physical processes are performed. Of particular interest is reports on actual problems of nuclear energetics and fuel cycle, on ill-intentioned use of fissile materials, efficiency of long-lived isotopes transmutation and spent fuel management [ru

  14. Conceptual research on reactor core physics for accelerator driven sub-critical reactor

    International Nuclear Information System (INIS)

    Zhao Zhixiang; Ding Dazhao; Liu Guisheng; Fan Sheng; Shen Qingbiao; Zhang Baocheng; Tian Ye

    2000-01-01

    The main properties of reactor core physics are analysed for accelerator driven sub-critical reactor. These properties include the breeding of fission nuclides, the condition of equilibrium, the accumulation of long-lived radioactive wastes, the effect from poison of fission products, as well as the thermal power output and the energy gain for sub-critical reactor. The comparison between thermal and fast system for main properties are carried out. The properties for a thermal-fast coupled system are also analysed

  15. Physical characteristics as determinants of adolescents' moral ...

    African Journals Online (AJOL)

    This study was designed to investigate the physical characteristics as determinants of adolescent moral behaviour in Ekiti State secondary schools. Two research hypotheses were formulated. The research design used was the descriptive of the survey type. The population for this study was made up of all adolescents in ...

  16. Some Physical Characteristics of Microcrystalline Cellulose ...

    African Journals Online (AJOL)

    Purpose: The microcrystalline cellulose is an important ingredient in pharmaceutical, food, cosmetic and other industries. This study aimed at evaluating the physical characteristics of microcrystalline cellulose (CP-MCC), obtained from the raw cotton of Cochlospermum planchonii. Methods: CP-MCC was obtained from the ...

  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. 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

  19. Reactor physics verification of the MCNP6 unstructured mesh capability

    Energy Technology Data Exchange (ETDEWEB)

    Burke, T. P. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States); Kiedrowski, B. C.; Martz, R. L. [X-Computational Physics Division, Monte Carlo Codes Group, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Martin, W. R. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109 (United States)

    2013-07-01

    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)

  20. Cold fusion reactors and new modern physics

    OpenAIRE

    Huang Zhenqiang Huang Yuxiang

    2013-01-01

    The author of the "modern physics classical particle quantization orbital motion model general solution", referred to as the “new modern physics” a book. “The nuclear force constraint inertial guidance cold nuclear fusion collides” patent of invention referred to as the “cold nuclear fusion reactor” detailed technical data. Now provide to you, hope you help spread and the mainstream of modern physics of academic and fusion engineering academic communication. We work together to promote the c...

  1. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    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. 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

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

    International Nuclear Information System (INIS)

    Briggs, J. Blair; Gulliford, Jim

    2014-01-01

    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.

  5. 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

  6. Achievements and future directions in the reactors physics and nuclear safety research

    International Nuclear Information System (INIS)

    Dumitrache, Ion

    2001-01-01

    A historical overlook is presented with respect to inception and development of reactor physics research and on the job training in Romania. First these activities were carried out at the Institute for Atomic Physics and Institute for Power Reactors (IRNE) in Bucharest and afterward at the Institute for Nuclear Technologies, later on transformed in the Institute of Nuclear Research at Pitesti. CYBER Computer installed at Pitesti allowed formation in as early as 1971 reactor specialists who worked out computer programs for neutron physics calculations. These specialists were able to assimilate the characteristic of CANDU 6 type reactor as well as the AECL methodology of simulating processes of CANDU reactor physics. At present four programs are under way. These are: 1. The nuclear reactor physics; 2. The nuclear facility safety; 3. Safety analyses for the transport and radioactive waste disposal; 4. Analyses for radiation shielding and biological protection. There are presented results of the work associated to the CANDU type reactor: 1. Adapting and improving the code system for neutron and thermohydraulic calculation for CANDU type reactor, as supplied by AECL; 2. The IRNE manual for CANDU reactor neutron designing; 3. Final sizing of shim rods of Cernavoda NPP Unit 2; 4. Tests and measurements of reactor physics at the Cernavoda NPP Unit 1 commissioning; 5. Simulation and independent analysis of thermosiphoning carried out at Cernavoda NPP Unit 1 commissioning; 6. Static and dynamical response of the detectors in the CANDU reactor core and their time evolution following the burnup in the neutron flux and their ageing effects; 7. PSA studies at Unit 1; 8. Safety analyses for the radioactive waste disposal at Saligny repository. Also, reported are the results of the work associated to the TRIGA reactor, as follows: 1. Flux measurements and neutron computations necessary in the reactor commissioning; 2. Cleaning up controversial issues relating to neutron flux

  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. Advances in Reactor physics, mathematics and computation. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

  9. 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

  10. Compatibility analysis of DUPIC fuel (Part II) - Reactor physics design and analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Choi, Hang Bok; Rhee, Bo Wook; Roh, Gyu Hong; Kim, Do Hun [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    The compatibility analysis of the DUPIC fuel in a CANDU reactor has been assessed. This study includes the fuel composition adjustment, comparison of lattice properties, performance analysis of reactivity devices, determination of regional over-power (ROP) trip setpoint, and uncertainty estimation of core performance parameters. For the DUPIC fuel composition adjustment, three options have been proposed, which can produce uniform neutronic characteristics of the DUPIC fuel. The lattice analysis has shown that the characteristics of the DUPIC fuel is compatible with those of natural uranium fuel. The reactivity devices of the CANDU-6 reactor maintain their functional requirements even for the DUPIC fuel system. The ROP analysis has shown that the trip setpoint is not sacrificed for the DUPIC fuel system owing to the power shape that enhances more thermal margin. The uncertainty analysis of the core performance parameter has shown that the uncertainty associated with the fuel composition variation is reduced appreciably, which is primarily due to the fuel composition adjustment and secondly the on-power refueling feature and spatial control function of the CANDU reactor. The reactor physics calculation has also shown that it is feasible to use spent PWR fuel directly in CANDU reactors without deteriorating the CANDU-6 core physics design requirements. 29 refs., 67 figs., 60 tabs. (Author)

  11. Project and characteristics of a 5MW experimental fast reactor

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1986-05-01

    Characteristics of a 5 MW experimental fast reactor are reported. The reactor is designed with emphasis on fuel and materials irradiation and uses fuel assemblies of a standard structure. The reference core consist of 37 fuel assemblies, each of which contains 19 pins of metallic Pu/Zr fuel. With a core height of 17.6 cm the core volume is 11.4 liter and the central fast (E >=100 KeV) flux is 0.9 x 10 15 n/cm 2 sec. In addition to twelve control rod assemblies with a total reactivity worth of 5.5% Δk, 42 assemblies for reactivity compensation are placed in the two rings outside the core. Replacing these assemblies with driver, blanket, or refletor-shield assemblies, large reactivities can be added to make the central assembly position available for test irradiations and to assure high levels of burnup of driver assemblies. (Author) [pt

  12. Nuclear characteristics of D-D fusion reactor blankets, (1)

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Ohta, Masao; Seki, Yasushi.

    1977-01-01

    Fusion reactors operating on the deuterium (D-D) cycle are considered promising for their freedom from tritium breeding in the blanket. In this paper, neutronic and photonic calculations are undertaken covering several blanket models of the D-D fusion reactor, using presently available data, with a view to comparing the nuclear characteristics of these models, in particular, the nuclear heating rates and their spatial distributions. Nine models are taken up for the study, embodying various combinations of coolant, blanket, structural and reflector materials. About 10 MeV is found to be a typical value for the total nuclear energy deposition per source neutron in the models considered here. The realization of high energy gain is contingent upon finding a favorable combination of blanket composition and configuration. The resulting implications on the thermal design aspect are briefly discussed. (auth.)

  13. 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)

  14. 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

  15. 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

  16. Physical and chemical characteristics of fibrous peat

    Science.gov (United States)

    Sutejo, Yulindasari; Saggaff, Anis; Rahayu, Wiwik; Hanafiah

    2017-11-01

    Banyuasin is one of the regency in South Sumatera which has an area of 200.000 Ha of peat land. Peat soil are characterized by high compressibility parameters and low initial shear strength. Block sampling method was used to obtain undisturbed sample. The results of this paper describe the characteristics of peat soil from physical and chemical testing. The physical and chemical characteristics of peat include water content (ω), specific gravity (Gs), Acidity (pH), unit weight (γ), and ignition loss tests. SEM and EDS test was done to determine the differences in fiber content and to analyze chemical elements of the specimen. The average results ω, Gs, and pH are 263.538 %, 1.847, and 3.353. Peat is classified in H4 (by Von Post). The results of organic content (OC), ash content (AC), and fiber content (FC) are found 78.693 %, 21.310 %, and 73.703 %. From the results of physical and chemical tests, the peat in Banyuasin is classified as fibrous peat. All the results of the characteristics and classification of fibrous peat compared with published data were close.

  17. Experimental and calculational works on characteristics of the Dalat Nuclear Research Reactor. Second edition

    International Nuclear Information System (INIS)

    Pham Ngoc Khoi; Nguyen Kim Dung

    2016-03-01

    Recognizing the significant value and necessity of publishing the scientific document of experimental and calculational works on the Dalat Nuclear Research Reactor (DNRR) physics and engineering for research, operation, training activities as well as for international scientific exchange, Vietnam Atomic Energy Agency (VAEA) and Vietnam Atomic Energy Institute have completed editing to publish the “Experimental and Calculational Works on Characteristics of THE DALAT NUCLEAR RESEARCH REACTOR” which consists of 26 typical papers representing the most important experimental and calculational results of the DNRR physics and engineering obtained during 30 years of operation and exploitation with the contribution of Vietnamese and former USSR’s experts, especially scientists and engineers working at the Reactor Center of the NRI

  18. 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

  19. Final Physics Report for the Engineering Test Reactor

    International Nuclear Information System (INIS)

    Wolfe, I. B.

    1956-01-01

    This report is a summary of the physics design work performed on the Engineering Test Reactor. The ETR presents computational difficulties not found in other reactors because of the large number of experimental holes in the core. The physics of the ETR depends strongly upon the contents of the in-core experimental facilities. In order to properly evaluate the reactor' taking into account the experiments in the core, multi-region, two-dimensional calculations are required. These calculations require the use of a large computer such as the Remington Rand Univac and are complex and expensive enough to warrant a five-stage program: 1. In the early stages of design, only preliminary two-dimensional calculations were performed .in order to obtain a rough idea of the general behavior of the reactor and its critical mass with tentative experiments in place. 2. A large amount of work was carried out in which the reactor was approximated as one with a uniform homogeneous core. With this model, detailed studies were carried out to investigate the feasibility and to obtain general design data on such points as the design and properties of the gray and black control rods, the design of the beryllium reflector, gamma and neutron heating, the use of burnable poisons, etc. In performing these calculations, use was made of the IBM 650 PROD code obtained from KAPL. 3. With stages 1 and 2 carried out, two-dimensional calculations of the core at start-up conditions were performed on the Univac computer. 4. Detailed two-dimensional calculations of the properties of the ETR with a proposed first set of experiments in place were carried out. 5. A series of nuclear tests were performed at the reactivity measurements facility at the MTR site in order to confirm the validity of the analytical techniques in physics analysis. In performing the two-dimensional Univac calculations, the MUG code developed by KAPL and the Cuthill code developed at the David Taylor Model Basin were utilized. In

  20. 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

  1. Response characteristics of reactor building on weathered soft rock ground

    International Nuclear Information System (INIS)

    Hirata, Kazuta; Tochigi, Hitoshi

    1991-01-01

    The purpose of this study is to investigate the seismic stability of nuclear power plants on layered soft bedrock grounds, focusing on the seismic response of reactor buildings. In this case, the soft bedrock grounds refer to the weathered soft bedrocks with several tens meter thickness overlaying hard bedrocks. Under this condition, there are two subjects regarding the estimation of the seismic response of reactor buildings. One is the estimation of the seismic response of surface ground, and another is the estimation of soil-structure interaction characteristics for the structures embedded in the layered grounds with low impedandce ratio between the surface ground and the bedrock. Paying attention to these subjects, many cases of seismic response analysis were carried out, and the following facts were clarified. In the soft rock grounds overlaying hard bedrocks, it was proved that the response acceleration was larger than the case of uniform hard bedrocks. A simplified sway and rocking model was proposed to consider soil-structure interaction. It was proved that the response of reactor buildings was small when the effect of embedment was considered. (K.I.)

  2. Reactor physical experimental program EROS in the frame of the molten salt applying reactor concepts development

    International Nuclear Information System (INIS)

    Hron, Miloslav; Kyncl, Jan; Mikisek, Miroslav

    2009-01-01

    After the relatively broad program of experimental activities, which have been involved in the complex R and D program for the Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept development in the Czech Republic, there has been a next stage (namely large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0 called EROS project) started, which will be focused to the experimental verification of the rector physical or neutronic properties of other types of reactor concepts applying molten salts in the role of liquid fuel and/or coolant. This tendency is based on the recently accepted decision of the MSR SSC of GIF to consider for further period of its activity two baseline concepts- fast neutron molten salt reactor non-moderated (FMSR-NM) as a long-term alternative to solid fuelled fast neutron reactors and simultaneously, advanced high temperature reactor (AHTR) with pebble bed type solid fuel cooled by liquid salts. There will be a brief description of the prepared and performed experimental programs in these directions (as well as the preliminary results obtained so far) introduced in the paper. (author)

  3. Physics design of an ultra-long pulsed tokamak reactor

    International Nuclear Information System (INIS)

    Ogawa, Y.; Inoue, N.; Wang, J.; Yamamoto, T.; Okano, K.

    1993-01-01

    A pulsed tokamak reactor driven only by inductive current drive has recently revived, because the non-inductive current drive efficiency seems to be too low to realize a steady-state tokamak reactor with sufficiently high energy gain Q. Essential problems in pulsed operation mode is considered to be material fatigue due to cyclic operation and expensive energy storage system to keep continuous electric output during a dwell time. To overcome these problems, we have proposed an ultra-long pulsed tokamak reactor called IDLT (abbr. Inductively operated Day-Long Tokamak), which has the major and minor radii of 10 m and 1.87 m, respectively, sufficiently to ensure the burning period of about ten hours. Here we discuss physical features of inductively operated tokamak plasmas, employing the similar constraints with ITER CDA design for engineering issues. (author) 9 refs., 2 figs., 1 tab

  4. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  5. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    International Nuclear Information System (INIS)

    Jordan, K. A.; Schubring, D.; Girardin, G.; Pautz, A.

    2013-01-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  6. Mixing Characteristics during Fuel Coolant Interaction under Reactor Submerged Conditions

    International Nuclear Information System (INIS)

    Hong, S. W.; Na, Y. S.; Hong, S. H.; Song, J. H.

    2014-01-01

    A molten material is injected into an interaction chamber by free gravitation fall. This type of fuel coolant interaction could happen to operating plants. However, the flooding of a reactor cavity is considered as SAM measures for new PWRs such as APR-1400 and AP1000 to assure the IVR of a core melt. In this case, a molten corium in a reactor is directly injected into water surrounding the reactor vessel without a free fall. KAERI has carried out fuel coolant interaction tests without a free fall using ZrO 2 and corium to simulate the reactor submerged conditions. There are four phases in a steam explosion. The first phase is a premixing phase. The premixing is described in the literature as follows: during penetration of melt into water, hydrodynamic instabilities, generated by the velocities and density differences as well as vapor production, induce fragmentation of the melt into particles; the particles fragment in turn into smaller particles until they reach a critical size such that the cohesive forces (surface tension) balance exactly the disruptive forces (inertial); and the molten core material temperature (>2500 K) is such that the mixing always occurs in the film boiling regime of the water: It is very important to qualify and quantify this phase because it gives the initial conditions for a steam explosion This paper mainly focuses on the observation of the premixing phase between a case with 1 m free fall and a case without a free fall to simulate submerged reactor condition. The premixing behavior between a 1m free fall case and reactor case submerged without a free fall is observed experimentally. The average velocity of the melt front passing through 1m water pool; - Case without a free fall: The average velocity of corium, 2.7m/s, is faster than ZrO 2 , 2.3m/s, in water. - Cases of with a 1 m free fall and without a free fall : The case without a free fall is about two times faster than a case with a 1 m free fall. Bubble characteristics; - Case

  7. Flow characteristics of Korea multi-purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heonil Kim; Hee Taek Chae; Byung Jin Jun; Ji Bok Lee [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-09-01

    The construction of Korea Multi-purpose Research Reactor (KMRR), a 30 MW{sub th} open-tank-in-pool type, is completed. Various thermal-hydraulic experiments have been conducted to verify the design characteristics of the KMRR. This paper describes the commissioning experiments to determine the flow distribution of KMRR core and the flow characteristics inside the chimney which stands on top of the core. The core flow is distributed to within {+-}6% of the average values, which is sufficiently flat in the sense that the design velocity in the fueled region is satisfied. The role of core bypass flow to confine the activated core coolant in the chimney structure is confirmed.

  8. 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

  9. Physical and technical aspects of lead cooled fast reactors safety

    International Nuclear Information System (INIS)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I.

    2001-01-01

    The safety analysis of lead-cooled fast reactors has been performed for the well-developed concept of BREST-OD-300 reactor. The most severe accidents have been considered. An ultimate design-basis accident has been defined as an event resulting from an external impact and involving a loss of leak-tightness of the lead circuit, loss of forced circulation of lead and loss of heat sink to the secondary circuit, failure of controls and of reactor scram with resultant insertion of total reactivity margin, etc. It was assumed in accident analysis that the protective feature available for accident mitigation was only reactivity feedback on the changes in the temperatures of the reactor core elements and coolant flow rate, and in some cases also actuation of passive protections of threshold action in response to low flow rate and high coolant temperature at the core outlet. It should be noted that the majority of the analyzed accidents could be overcame even without initiation of the above protections. It has been demonstrated that a combination of inherent properties of lead coolant, nitride fuel, physical and design features of fast reactors will ensure natural safety of BREST and are instrumental for avoiding by a deterministic approach the accidents associated with a significant release of radioactivity and requiring evacuation of people in any credible initiating event and a combination of events. (author)

  10. Study and application of boiling water reactor jet pump characteristic

    International Nuclear Information System (INIS)

    Liao Lihyih

    1992-01-01

    RELAP5/MOD2 is an advanced thermal-hydraulic computer code used to analyze plant response to postulated transient and loss-of-coolant accidents in light water nuclear reactors. Since this computer code was originally developed for pressurized water reactor transient analysis, some of its capabilities are questioned when the methods are applied to a boiling water reactor. One of the areas which requires careful assessment is the jet pump model. In this paper, the jet pump models of RELAP5/MOD2, RETRAN-02/MOD3, and RELAP4/MOD3 are compared. From an investigation of the momentum equations, it is found that the jet pump models of these codes are not exactly the same. However, the effects of the jet pump models on the M-N characteristic curve are negligible. In this study, it is found that the relationship between the flow ratio, M, and the head ratio, N, is uniquely determined for a given jet pump geometry provided that the wall friction and gravitational head are neglected. In other words, under the given assumptions, the M-N characteristic curve will not change with power, level, recirculation pump speed or loop flow rate. When the effects of wall friction and gravitational head are included, the shape of the M-N curve will change. For certain conditions, the slope of the M-N curve can even change from negative to positive. The changes in the M-N curve caused by the separate effects of the wall friction and gravitational head will be presented. Sensitivity studies on the drive flow nozzle form loss coefficients, K d , the suction flow junction form loss coefficients, K s , the diffuser form loss coefficient, K c , and the ratio of different flow areas in the jet pump are performed. Finally, useful guidelines will be presented for plants without a plant specific M-N curve. (orig.)

  11. Neutron physics computation of CERCA fuel elements for Maria Reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.J.; Kulikowska, T.; Marcinkowska, Z.

    2008-01-01

    Neutron physics parameters of CERCA design fuel elements were calculated in the framework of the RERTR (Reduced Enrichment for Research and Test Reactors) program for Maria reactor. The analysis comprises burnup of experimental CERCA design fuel elements for 4 cycles in Maria Reactor To predict the behavior of the mixed core the differences between the CERCA fuel (485 g U-235 as U 3 Si 2 , 5 fuel tubes, low enrichment 19.75 % - LEU) and the presently used MR-6 fuel (430 g as UO 2 , 6 fuel tubes, high enrichment 36 % - HEU) had to be taken into account. The basic tool used in neutron-physics analysis of Maria reactor is program REBUS using in its dedicated libraries of effective microscopic cross sections. The cross sections were prepared using WIMS-ANL code, taking into account the actual structure, temperature and material composition of the fuel elements required preparation of new libraries.The problem is described in the first part of the present paper. In the second part the applicability of the new library is shown on the basis of the fuel core computational analysis. (author)

  12. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  13. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    International Nuclear Information System (INIS)

    Heeger, Karsten M.

    2014-01-01

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta . Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  14. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

  15. 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)

  16. Flux-limited diffusion coefficients in reactor physics applications

    International Nuclear Information System (INIS)

    Pounders, J.; Rahnema, F.; Szilard, R.

    2007-01-01

    Flux-limited diffusion theory has been successfully applied to problems in radiative transfer and radiation hydrodynamics, but its relevance to reactor physics has not yet been explored. The current investigation compares the performance of a flux-limited diffusion coefficient against the traditionally defined transport cross section. A one-dimensional BWR benchmark problem is examined at both the assembly and full-core level with varying degrees of heterogeneity. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Racz, A [ed.

    1996-12-31

    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.).

  18. 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.)

  19. 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.

  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

    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.

  2. 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)

  3. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  4. Passive safety design characteristics of the KALIMER-600 burner reactor

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Cho, Chung-Ho; Ha, Ki-Seok; Kim, Sang-Ji

    2009-01-01

    The Korea Atomic Energy Research Institute (KAERI) has recently studied several burner core designs for a transuranics (TRU) transmutation based on the breakeven core geometry of KALIMER-600. The KALIMER-600 is a net electrical rating of 600MWe, sodium-cooled, metallic-fueled, pool-type reactor. For the burner core concept selected for the present analysis, the smearing fractions of the fuel rods in three fuel zones are changed while maintaining the cladding outer diameter and cladding thickness. The resulting fuel slug smearing fractions of the inner, middle, and outer core zones are 36%, 40%, and 48%, respectively. The TRU conversion ratio is 0.57 and the TRU enrichment of the driver fuel is set to 30.0 w/o because of the current practical limitation of the U-TRU-10%Zr metal fuel database. The purpose of this paper is to evaluate the safety performance characteristics provided by the passive safety design features in the KALIMER-600 burner reactor by using a system-wide safety analysis code. The present scoping analysis focuses on an assessment of the enhanced safety design features that provide passive and self-regulating responses to transient conditions and an evaluation of the safety margin during unprotected overpower, unprotected loss of flow, and unprotected loss of heat sink events. The analysis results show that the KALIMER-600 burner reactor provides larger safety margins with respect to the sodium boiling, fuel rod integrity, and structural integrity. The overall inherent safety can be enhanced by accounting for the reactivity feedback mechanisms in the design process. (author)

  5. Current status of the reactor physics code WIMS and recent developments

    International Nuclear Information System (INIS)

    Lindley, B.A.; Hosking, J.G.; Smith, P.J.; Powney, D.J.; Tollit, B.S.; Newton, T.D.; Perry, R.; Ware, T.C.; Smith, P.N.

    2017-01-01

    Highlights: • The current status of the WIMS reactor physics code is presented. • Applications range from 2D lattice calculations up to 3D whole core geometries. • Gamma transport and thermal-hydraulic feedback models added. • Calculations methodologies described for several Gen II, III and IV reactor types. - Abstract: The WIMS modular reactor physics code has been under continuous development for over fifty years. This paper discusses the current status of WIMS and recent developments, in particular developments to the resonance shielding methodology and 3D transport solvers. Traditionally, WIMS is used to perform 2D lattice calculations, typically to generate homogenized reactor physics parameters for a whole core code such as PANTHER. However, with increasing computational resources there has been a growing trend for performing transport calculations on larger problems, up to and including 3D full core models. To this end, a number of the WIMS modules have been parallelised to allow efficient performance for whole core calculations, and WIMS includes a 3D method of characteristics solver with reflective and once-through tracking methods, which can be used to analyse problems of varying size and complexity. A time-dependent flux solver has been incorporated and thermal-hydraulic modelling capability is also being added to allow steady-state and transient coupled calculations to be performed. WIMS has been validated against a range of experimental data and other codes, in particular for water and graphite moderated thermal reactors. Future developments will include improved parallelization, enhancing the thermal-hydraulic feedback models and validating the WIMS/PANTHER code system for BWRs and fast reactors.

  6. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  7. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  8. Fusion reactors: physics and technology. Annual progress report

    International Nuclear Information System (INIS)

    Conn, R.W.

    1983-08-01

    Fusion reactors are designed to operate at full power and generally at steady state. Yet experience shows the load variations, licensing constraints, and frequent sub-system failures often require a plant to operate at fractions of rated power. The aim of this study has been to assess the technology problems and design implications of startup and fractional power operation on fusion reactors. The focus of attention has been tandem mirror reactors (TMR) and we have concentrated on the plasma and blanket engineering for startup and fractional power operation. In this report, we first discuss overall problems of startup, shutdown and staged power operation and their influence on TMR design. We then present a detailed discussion of the plasma physics associated with TMR startup and various means of achieving staged power operation. We then turn to the issue of instrumentation and safety controls for fusion reactors. Finally we discuss the limits on transient power variations during startup and shutdown of Li 17 Pb 83 cooled blankets

  9. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Bouveret, F.

    2001-01-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  10. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    International Nuclear Information System (INIS)

    Ebrahimia, Mahsa; Suha, Kune Y.; Eghbalic, Rahman; Jahan, Farzaneh Asadi malek

    2012-01-01

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran

  11. Comparison of some characteristics of aerobic granules and sludge flocs from sequencing batch reactors.

    Science.gov (United States)

    Li, J; Garny, K; Neu, T; He, M; Lindenblatt, C; Horn, H

    2007-01-01

    Physical, chemical and biological characteristics were investigated for aerobic granules and sludge flocs from three laboratory-scale sequencing batch reactors (SBRs). One reactor was operated as normal SBR (N-SBR) and two reactors were operated as granular SBRs (G-SBR1 and G-SBR2). G-SBR1 was inoculated with activated sludge and G-SBR2 with granules from the municipal wastewater plant in Garching (Germany). The following major parameters and functions were measured and compared between the three reactors: morphology, settling velocity, specific gravity (SG), sludge volume index (SVI), specific oxygen uptake rate (SOUR), distribution of the volume fraction of extracellular polymeric substances (EPS) and bacteria, organic carbon and nitrogen removal. Compared with sludge flocs, granular sludge had excellent settling properties, good solid-liquid separation, high biomass concentration, simultaneous nitrification and denitrification. Aerobic granular sludge does not have a higher microbial activity and there are some problems including higher effluent suspended solids, lower ratio of VSS/SS and no nitrification at the beginning of cultivation. Measurement with CLSM and additional image analysis showed that EPS glycoconjugates build one main fraction inside the granules. The aerobic granules from G-SBR1 prove to be heavier, smaller and have a higher microbial activity compared with G-SBR2. Furthermore, the granules were more compact, with lower SVI and less filamentous bacteria.

  12. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

    2012-10-15

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

  13. 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

  14. 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)

  15. Teachers' and Students' Perceptions of Effective Physics Teacher Characteristics

    Science.gov (United States)

    Korur, Fikret; Eryilmaz, Ali

    2012-01-01

    Problem Statement: What do teachers and students in Turkey perceive as the common characteristics of effective physics teachers? Purpose of Study: The first aim was to investigate the common characteristics of effective physics teachers by asking students and teachers about the effects of teacher characteristics on student physics achievement and…

  16. A Study on thermal-hydraulic characteristics of the coolant materials for the transmutation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; You, Young Woo; Cho, Jae Seon; Kim, Ju Youl; Kim, Do Hyoung; Kim, Yoon Ik; Yang, Hui Chang [Seoul National University, Taejon (Korea)

    1998-03-01

    The objective of this study is to provide the direction of transmutation reactor design in terms of thermal hydraulics especially through the analysis of thermal hydraulic characteristics of various candidate materials for the transmutation reactor coolant. In this study, the characteristics of coolant materials used in current nuclear power plants and candidate materials for transmutation reactor are analyzed and compared. To evaluate the thermal hydraulic characteristics, the preliminary thermal-hydraulic calculation is performed for the candidate coolant materials of transmutation reactor. An analysis of thermal-hydraulic characteristics of transmutation reactor. An analysis of thermal-hydraulic characteristics of Sodium, Lead, Lead-Bismuth, and Lead-Lithium among the liquid metals considered as the coolant of transmutation reactor is performed by using computational fluid dynamics code FLUENT, and SIMPLER algorithm. (author). 50 refs., 40 figs., 30 tabs.

  17. 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

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

    International Nuclear Information System (INIS)

    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

  19. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Haas, J.B.M. de; Klippel, H.T.; Hogenbirk, A.; Oppe, J.; Sciolla, C.M.; Stad, R.C.L. van der; Zhang, B.C.

    1997-06-01

    As part of the activities within the framework of the development of INCOGEN, a 'Dutch' conceptual design of a smaller HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRs, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (orig.)

  20. Reactor physics calculations on the Dutch small HTR concept

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Hass, J.B.M. De; Klippel, H.Th.; Hogenbirk, A.; Oppe, J.; Sciolla, C.; Stad, R.C.L. Van Der; Zhang, B.C.

    1997-01-01

    As part of the activities within the framework of the development of INCOGEN, a ''Dutch'' conceptual design of a small HTR, the ECN reactor physics code system has been extended with the capability to perform combined neutronics and thermal hydraulics steady-state, burnup and transient core calculations on pebble-bed type HTRS, by joining the general purpose reactor code PANTHER and the HTR thermal hydraulics code THERMIX/DIREKT in the PANTHERMIX code combination. The validation of the ECN code system for HTR applications is still in progress, but some promising first calculation results on unit cell and whole core geometries are presented, which indicate that the extended ECN code system is quite suitable for performing the pebble-bed HTR core calculations, required in the INCOGEN core design and optimization process. (author)

  1. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  2. Physical characteristics of cinnamon oil microcapsule

    Science.gov (United States)

    Hermanto, R. F.; Khasanah, L. U.; Kawiji; Atmaka, W.; Manuhara, G. J.; Utami, R.

    2016-02-01

    Cinnamon (Cinnamomum burmanii) oil products can be obtained from the bark by steam distillation. Essential oils are susceptible to high temperatures, oxidation, UV light, and humidity. Microencapsulation may change essential oils into powder, protect the sensitive core material and reduce the amount of flavor which lost during storage. In the microencapsulation, one of the important factors is the type of coating agent. The objective of this work was to characterize the cinnamon oil microcapsule. Ratio variations of coating agent maltodextrin and gum arabic were (1:0); (0:1); (1:1); (2:3). Physical characteristics such as water content, solubility, bulk density, surface oil, and microencapsulation efficiency of samples were investigated. Results showed that the ratio variations of the coating agent significantly affected the water content, bulk density, surface oil and microencapsulation efficiency but significantly affected the water solubility. Characteristics of selected microcapsule were 6.13% water content; 96.33% solubility; 0.46 g/cm3 bulk density; 2.68% surface oil; 70.68% microencapsulation efficiency and microstructures were rather good.

  3. Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

    International Nuclear Information System (INIS)

    Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

    2008-01-01

    TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

  4. 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

  5. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Science.gov (United States)

    2010-01-01

    ... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... perform their duties. (6) Prior to entry into a material access area, packages shall be searched for...

  6. Characteristics of irradiation creep in the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Coghlan, W.A.; Mansur, L.K.

    1981-01-01

    A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

  7. Design characteristics of zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Popovic, D.; Nikolic, D.; Antic, D.; Zavaljevski, N.

    1987-01-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  8. Benchmark problem suite for reactor physics study of LWR next generation fuels

    International Nuclear Information System (INIS)

    Yamamoto, Akio; Ikehara, Tadashi; Ito, Takuya; Saji, Etsuro

    2002-01-01

    This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70 GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO 2 and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management. (author)

  9. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    Energy Technology Data Exchange (ETDEWEB)

    Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Blaise, P. [CEA, DEN, DER, SPEX Experimental Programs Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

  10. 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.)

  11. Development of a new physics data library for the SRS reactors

    International Nuclear Information System (INIS)

    Niemer, K.A.

    1993-01-01

    The Savannah River Site (SRS) reactors have historically operated at power levels of -2500 MW; thus, previous reactor physics data libraries were created based on that constant power. However, as a result of recent lower power operation, the existing physics data libraries are no longer adequate. Therefore, a new power-dependent physics library was needed to model the reactor at different power levels. The design and development of a new power-dependent physics data library is discussed in this paper

  12. 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)

  13. Basic experiments of reactor physics using the critical assembly TCA

    International Nuclear Information System (INIS)

    Obara, Toru; Igashira, Masayuki; Sekimoto, Hiroshi; Nakajima, Ken; Suzaki, Takenori.

    1994-02-01

    This report is based on lectures given to graduate students of Tokyo Institute of Technology. It covers educational experiments conducted with the Tank-Type Critical Assembly (TCA) at Japan Atomic Energy Research Institute in July, 1993. During this period, the following basic experiments on reactor physics were performed: (1) Critical approach experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, (5) Measurement of safety sheet worth by the rod drop method. The principle of experiments, experimental procedure, and analysis of results are described in this report. (author)

  14. 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

  15. The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

    International Nuclear Information System (INIS)

    Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

    2010-01-01

    LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

  16. 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)

  17. 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)

  18. Transient thermal characteristics of a core channel in a molten salt reactor

    International Nuclear Information System (INIS)

    Sakashita, H.; Ishiguro, R.; Sugiyama, K.

    1987-01-01

    The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

  19. Research about reactor operator's personability characteristics and performance

    Energy Technology Data Exchange (ETDEWEB)

    Wei Li; He Xuhong; Zhao Bingquan [Tsinghua Univ., Institute of Nuclear Energy Technology, Beijing (China)

    2003-03-01

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  20. 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.

  1. Passive devices of a reactor stop: classification of the characteristics and estimation of perfection degree

    International Nuclear Information System (INIS)

    Portyanoj, A.G.; Serdun', E.N.; Sorokin, A.P.; Egorov, V.S.; Shkarovskij, D.A.

    1998-01-01

    The perspective direction in NPP safety improvement connected with development of passive devices for nuclear reactor emergency shutdown (PDRS) is discussed. More than hundred devices which can fulfil the PDRS functions are suggested nowadays. The analysis of PDRS designing status as applicable for the fast reactors in the main which are based on the physical effect used in an element sensitive to temperature is made. The complex consisting of nine general characteristics including passive character, thresholdness, forces generation, inertia, multichannel design, stability towards operational factors, safety at failures, simplicity and visualisation, development conditions, is suggested for estimation of the quality of PDRS of different types. Basing on expert assessments realized using the complex of general characteristics it is shown that the types of PDRS may be separated into following three groups: linear expansion of solid bodies and thermoelectric ones (K ≅ 0.45); magnet ones with shape memory effect, liquid volume expansion (K ≅ 0.6); fusing ones (K ≅ 0.7). The conclusion is made that PDRS on the basis of fusing devices of the sulphon type with liofobic capillary-porous working body most completely satisfy the complex of general characteristics considered

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-02-15

    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.

  4. 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)

  5. International Thermonuclear Experimental Reactor: Physics issues, capabilities and physics program plans

    International Nuclear Information System (INIS)

    Wesley, J.C.

    1997-01-01

    Present status and understanding of the principal plasma-performance determining physics issues that affect the physics design and operational capabilities of the International Thermonuclear Experimental Reactor (ITER) [ITER EDA Agreement and Protocol 2 (International Atomic Energy Agency, Vienna, 1994)] are presented. Emphasis is placed on the five major physics-basis issues emdash energy confinement, beta limit, density limit, impurity dilution and radiation loss, and the feasibility of obtaining partial-detached divertor operation emdash that directly affect projections of ITER fusion power and burn duration performance. A summary of these projections is presented and the effect of uncertainties in the physics-basis issues is examined. ITER capabilities for experimental flexibility and plasma-performance optimization are also described, and how these capabilities may enter into the ITER physics program plan is discussed. copyright 1997 American Institute of Physics

  6. Safety Analysis for Medium/Small Size Integral Reactor: Evaluation of Safety Characteristics for Small and Medium Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hho jung; Seul, K W; Ahn, S K; Bang, Y S; Park, D G; Kim, B K; Kim, W S; Lee, J H; Kim, W K; Shim, T M; Choi, H S; Ahn, H J; Jung, D W; Kim, G I; Park, Y M; Lee, Y J [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1997-07-01

    The Small and medium integral reactor is developed to be utilized for non-electric areas such as district heating and steam production for desalination and other industrial purposes, and then these applications may typically imply a closeness between the reactor and the user. It requires the reactor to be designed with the adoption of special functional and inherent safety features to ensure and promote a high level of safety and reliability, in comparison with the existing nuclear power plants. The objective of the present study is to establish the bases for the development of regulatory requirements and technical guides to address the special safety characteristics of the small and medium integral reactor. In addition, the study aims to identify and to propose resolutions to the possible safety concerns in the design of the small and medium integral reactor. 34 refs., 20 tabs. (author)

  7. Methods for reactor physics calculations for control rods in fast reactors

    International Nuclear Information System (INIS)

    Grimstone, M.J.; Rowlands, J.L.

    1988-12-01

    The IAEA Specialists' Meeting on ''Methods for Reactor Physics Calculations for Control Rods in Fast Reactors'' was held in Winfrith, United Kingdom, on 6-8 December, 1988. The meeting was attended by 23 participants from nine countries. The purpose of the meeting was to review the current calculational methods and their accuracy as assessed by theoretical studies and comparisons with measurements, and then to identify the requirements for improved methods or additional studies and comparisons. The control rod properties or effects to be considered were their reactivity worths, their effect on the power distribution through the core, and the reaction rates and energy deposition both within and adjacent to the rods. The meeting was divided into five sessions, in the first of which each national delegation presented a brief overview of their programme of work on calculational methods for fast reactor control rods. In the next three sessions a total of seventeen papers were presented describing calculational methods and assessments of their accuracy. The final session was a discussion to draw conclusions regarding the current status of methods and the further developments and validation work required. A separate abstract was prepared for each of the 23 papers presented at the meeting. Refs, figs and tabs

  8. The next generation of power reactors - safety characteristics

    International Nuclear Information System (INIS)

    Modro, S.M.

    1995-01-01

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

  9. Start up physics tests of units 5 and 6 (WWER 1000) at Kozloduy NPP by comparison with the calculated neutron physics characteristics

    International Nuclear Information System (INIS)

    Antov, A.; Stoyanova, I.

    2008-01-01

    In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests, which are performed after each refuelling provides assurance that the plant can be operated as designed. The calculated neutron-physics characteristics values of Kozloduy NPP Unit 5 and Unit 6 (WWER 1000) obtained by the computer code package KASKAD are compared with the obtained results during the start up physics tests. The core fuel loading consists of 163 fuel assemblies (FAs). The calculated values are given according to actual experimental conditions of the reactor core during start up physics tests. The report includes comparisons between calculation results by code package KASKAD (BIPR7A) and experimental data values of main neutron-physics characteristics during start up physics tests in selected recent cycles of Kozloduy NPP Unit 5 and Unit 6. (authors)

  10. Start up physics tests of Units 5 and 6 (WWER-1000) at Kozloduy NPP by comparison with the calculated neutron physics characteristics

    International Nuclear Information System (INIS)

    Stoyanova, I.; Antov, A.

    2007-01-01

    In conjunction with each refuelling shutdown of the reactor core, nuclear design calculations are performed to ensure that the reactor physics characteristics of the new core will be consistent with the safety limits. Prior to return to normal operation, a physics test program is required to determine if the operating characteristics of the core are consistent with the design predictions and to ensure that the core can be operated as designed. Successful completion of the physics test program is demonstrated when the test results agree with the predicted results within predetermined test criteria. Successful completion of the physics test program and successful completion of other tests which are performed after each refuelling provides assurance that the plant can be operated as designed. The calculated neutron-physics characteristics values of Kozloduy NPP Unit 5 and Unit 6 (WWER 1000) obtained by the computer code package KASKAD are compared with the obtained results during the start up physics tests. The core fuel loading consists of 163 fuel assemblies. The calculated values are given according to actual experimental conditions of the reactor core during start up physics tests. The report includes comparisons between calculation results by code package KASKAD (BIPR7A) and experimental data values of main neutron-physics characteristics during start up physics tests in selected recent cycles of Kozloduy NPP Unit 5 and Unit 6 (Authors)

  11. 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

  12. 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

  13. Gas-liquid reactor / separator: dynamics and operability characteristics

    NARCIS (Netherlands)

    Ranade, V.; Kuipers, J.A.M.; Versteeg, Geert

    1999-01-01

    A comprehensive mathematical model is developed to simulate gas¿liquid reactor in which both, reactants as well as products enter or leave the reactor in gas phase while the reactions take place in liquid phase. A case of first-order reaction (isothermal) was investigated in detail using the dynamic

  14. Potential market and characteristics of low-temperature reactors

    International Nuclear Information System (INIS)

    Lerouge, B.

    1975-01-01

    The low-temperature (100 to 200 deg C) heat market for industrial applications and district heating is very important. Two main studies have been developed: a swimming pool reactor delivering water at 110 deg C and a prestressed concrete vessel reactor delivering water at 200 deg C [fr

  15. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

    International Nuclear Information System (INIS)

    Bess, John D.; Gullifor, Jim

    2015-03-01

    The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project 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 IRPhE Project 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 evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, 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, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, 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 nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

  16. Physically - engineering problems of the Salaspils Nuclear reactor: Solutions and their topicality

    International Nuclear Information System (INIS)

    Mozgirs, Z.V.

    2005-01-01

    The paper generalizes technical solutions of physically-engineering problems of the Salaspils nuclear research reactor, experience of its modernization and exploitation. New equipment and the related technical solutions have been tested at the Salaspils reactor during its operation time and are now recommended for further use at nuclear reactors. (author)

  17. 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)

  18. Educational reactor-physics experiments with the critical assembly TCA

    International Nuclear Information System (INIS)

    Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki; 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)

  19. SILOETTE, a training centre for reactor physics at the Nuclear Research Centre of Grenoble

    International Nuclear Information System (INIS)

    Destot, M.

    1983-10-01

    The Reactor Department of Grenoble has created, based on Siloette, an activity of training in reactor physics, wich is running since 1975 to meet the important needs generated by the development of electronuclear power stations. Its essential goal is to provide an initiation to the basic physical phenomena which determine the operation of the reactors. For that purpose, a rather comprehensive program of practical works on reactor (SILOETTE) and on nuclear power station simulators (PWR, UNGG) is proposed besides lectures and conferences, general and specialized teaching on the reactor operation principle, kinetics, dynamics and thermics

  20. Characteristics of Flameless Combustion in 3D Highly Porous Reactors under Diesel Injection Conditions

    Directory of Open Access Journals (Sweden)

    M. Weclas

    2013-01-01

    Full Text Available The heat release process in a free volume combustion chamber and in porous reactors has been analyzed under Diesel engine-like conditions. The process has been investigated in a wide range of initial pressures and temperatures simulating engine conditions at the moment when fuel injection starts. The resulting pressure history in both porous reactors and in free volumes significantly depends on the initial pressure and temperature. At lower initial temperatures, the process in porous reactors is accelerated. Combustion in a porous reactor is characterized by heat accumulation in the solid phase of the porous structure and results in reduced pressure peaks and lowered combustion temperature. This depends on reactor heat capacity, pore density, specific surface area, pore structure, and heat transport properties. Characteristic modes of a heat release process in a two-dimensional field of initial pressure and temperature have been selected. There are three characteristic regions represented by a single- and multistep oxidation process (with two or three slopes in the reaction curve and characteristic delay time distribution has been selected in five characteristic ranges. There is a clear qualitative similarity of characteristic modes of the heat release process in a free volume and in porous reactors. A quantitative influence of porous reactor features (heat capacity, pore density, pore structure, specific surface area, and fuel distribution in the reactor volume has been clearly indicated.

  1. 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

  2. Validation of the VTT's reactor physics code system

    International Nuclear Information System (INIS)

    Tanskanen, A.

    1998-01-01

    At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

  3. Noise analysis of the Dodewaard boiling water reactor: characteristics and time history

    International Nuclear Information System (INIS)

    Veer, J.H.C. v.d.; Kema, N.V.

    1982-01-01

    Reactor noise measurements have been performed in the Dodewaard BWR since the eighth fuel cycle (1978). Analysis of the noise characteristics of the ex-core neutron detectors are reported. As a result characteristics of the global component of the boiling noise and the influence of oscillatory effects in reactor pressure control and steam flow rate are described. The influence of power feedback effects on the detection of global noise at low frequencies is given using point kinetic reactor theory. Results are reported on the behaviour of the neutron noise characteristics during one fuel cycle and on the behaviour from fuel cycle 8 to 11. (author)

  4. 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

  5. Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

    2011-01-01

    Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

  6. The past, present, and future of test and research reactor physics

    International Nuclear Information System (INIS)

    Ryskamp, J.M.

    1992-01-01

    Reactor physics calculations have been performed on research reactors since the first one was built 50 yr ago under the University of Chicago stadium. Since then, reactor physics calculations have evolved from Fermi-age theory calculations performed with slide rules to three-dimensional, continuous-energy, coupled neutron-photon Monte Carlo computations performed with supercomputers and workstations. Such enormous progress in reactor physics leads us to believe that the next 50 year will be just as exciting. This paper reviews this transition from the past to the future

  7. Physical characteristics of the Laccadive sea (Lakshadweep)

    Digital Repository Service at National Institute of Oceanography (India)

    Varkey, M.J.; Das, V.K.; RamaRaju, D.V.

    and current. Considering the physical behaviour, the coastal waters of this highly industrialised region were found to possess high instantaneous dilution capacity. However the spatial and temporal variability of the parameters calls for better understanding...

  8. Breeding characteristics analysis of a commercial fast reactor cooled with sodium liquid

    International Nuclear Information System (INIS)

    Kosaka, N.; Shigehiro, A.

    1982-01-01

    The fast reactor breeding characteristics and its safety is analysed. As reference, for a preliminar analysis, the specifications of Super-Phenix, reactor french of 1200 MWe, are used, varying some parameters after aiming to verify its effects on duplication time. (E.G.) [pt

  9. 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

  10. Thermohydraulic characteristics under some transient conditions of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang; Khang, Ngo Phu; An, Tran Khac; Nghiem, Huynh Ton [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Some experimental and theoretical thermal hydraulic characteristics of the Dalat Nuclear Research Reactor are presented, together with some general assessments, from a thermal hydraulic point of view, of its safety under transient conditions. (author). 3 refs., 9 figs., 7 tabs.

  11. Effect of Biochar on Soil Physical Characteristics

    DEFF Research Database (Denmark)

    Sun, Zhencai; Møldrup, Per; Vendelboe, Anders Lindblad

    Biochar addition to agricultural soil has been reported to reduce climate gas emission, as well as improve soil fertility and crop productivity. Little, however, is known about biochar effects on soil structural characteristics. This study investigates if biochar-application changes soil structural...... characteristics, as indicated from water retention and gas transport measurements on intact soil samples. Soil was sampled from a field experiment on a sandy loam with four control plots (C) without biochar and four plots (B) with incorporated biochar at a rate of 20 tons per hectare (plot size, 6 x 8 m). The C...... and B plots were placed in a mixed sequence (C-B-C-B-C-B-C-B) and at the same time the eight plots formed a natural pH gradient ranging from pH 7.7 to 6.3. We determined bulk density, saturated hydraulic conductivity (K-sat), soil water retention characteristics, soil-air permeability, and soil...

  12. Constant physics and characteristics of fundamental constant

    International Nuclear Information System (INIS)

    Tarrach, R.

    1998-01-01

    We present some evidence which supports a surprising physical interpretation of the fundamental constants. First, we relate two of them through the renormalization group. This leaves as many fundamental constants as base units. Second, we introduce and a dimensional system of units without fundamental constants. Third, and most important, we find, while interpreting the units of the a dimensional system, that is all cases accessible to experimentation the fundamental constants indicate either discretization at small values or boundedness at large values of the corresponding physical quantity. (Author) 12 refs

  13. IRPhE - International Reactor Physics Experiments database

    International Nuclear Information System (INIS)

    Sartori, E.

    2004-01-01

    The OECD/NEA Nuclear Science Committee (NSC) has identified the need to establish international databases containing all the important experiments that are available for sharing among the specialists and has set up or sponsored specific activities to achieve this. The aim is to preserve them in an agreed standard format in computer accessible form, to use them for international activities involving validation of current and new calculational schemes including computer codes and nuclear data libraries, for assessing uncertainties, confidence bounds and safety margins, and to record measurement methods and techniques. It is a significant saving results from disseminating a standard benchmark set to be used worldwide. A framework for professionals that use the standard benchmark set to validate and verify modeling codes and data for radiation transport, criticality safety and reactor physics applications guarantees a comparative set of analyses. It represents also a good basis for pinpointing important gaps and where efforts should be concentrated and ensures knowledge and competence preservation, management and transfer in nuclear science and engineering. A large number of experimentalists, physicists, evaluators, modelers have devoted large amounts of their efforts and competencies to produce the data on which the methods we are using today are based. These data are far from having been exploited fully for the different nuclear and radiation technologies. This wealth of information needs to be preserved in a form more easily exploitable by modern information technology and for use in connection with novel and refined computational models with limitations of the past removed. These data will form the basis for the studies of more advanced nuclear technology, will be instrumental in identifying areas where there is a lack of knowledge and thus provide support to justifying new experiments that would reduce design uncertainties and consequently costs. Improvement of

  14. Small reactor technical and design characteristics proposed for Indonesia

    International Nuclear Information System (INIS)

    Nurdin, M.

    1992-01-01

    A Team for Small Nuclear Electricity Reactor has been formed in Indonesia since June 1990. It is responsible for assessment and design of a small reactor for electricity and/or sea-water desalination. This concept may become a good alternative for power-plants for small islands and for isolated areas in Indonesia, the system should function economically and environmentally sound. In addition to existing concepts, this presentation deals with modifications proposed in improving reliability and safety of reactor operation. For the size of 200 MWth or more (80 MWe or more), the possibility of designing an internal auxiliary heat removal system is discussed, hence there are two separate heat sinks for the core. Future development works for this concept should be directed in expanding their spectrum of utilization and their contribution to the national energy needs. (author). 7 refs., 4 tabs

  15. [Rapid startup and nitrogen removal characteristic of anaerobic ammonium oxidation reactor in packed bed biofilm reactor with suspended carrier].

    Science.gov (United States)

    Chen, Sheng; Sun, De-zhi; Yu, Guang-lu

    2010-03-01

    Packed bed biofilm reactor with suspended carrier was used to cultivate ANAMMOX bacteria with sludge inoculums from WWTP secondary settler. The startup of ANAMMOX reactor was comparatively studied using high nitrogen loading method and low nitrogen loading method with aerobically biofilmed on the carrier, and the nitrogen removal characteristic was further investigated. The results showed that the reactor could be started up successfully within 90 days using low nitrogen loading method, the removal efficiencies of ammonium and nitrite were nearly 100% and the TN removal efficiencywas over 75% , however, the high nitrogen loading method was proved unsuccessfully for startup of ANAMMOX reactor probably because of the inhibition effect of high concentration of ammonium and nitrite. The pH value of effluent was slightly higher than the influent and the pH value can be used as an indicator for the process of ANAMMOX reaction. The packed bed ANAMMOX reactor with suspended carrier showed good characteristics of high nitrogen loading and high removal efficiency, 100% of removal efficiency could be achieved when the influent ammonium and nitrite concentration was lower than 800 mg/L.

  16. Anthropometric and physical fitness characteristics of female ...

    African Journals Online (AJOL)

    These included 24 university-level, 17 provincial-level, and 14 national team players. Anthropometric data included age, stature, body mass, sum of seven skinfolds and percentage body fat. The fitness characteristics included flexibility, strength, explosive power, muscular endurance, agility, speed and aerobic endurance.

  17. Hybrid Reduced Order Modeling Algorithms for Reactor Physics Calculations

    Science.gov (United States)

    Bang, Youngsuk

    hybrid ROM algorithms which can be readily integrated into existing methods and offer higher computational efficiency and defendable accuracy of the reduced models. For example, the snapshots ROM algorithm is hybridized with the range finding algorithm to render reduction in the state space, e.g. the flux in reactor calculations. In another implementation, the perturbation theory used to calculate first order derivatives of responses with respect to parameters is hybridized with a forward sensitivity analysis approach to render reduction in the parameter space. Reduction at the state and parameter spaces can be combined to render further reduction at the interface between different physics codes in a multi-physics model with the accuracy quantified in a similar manner to the single physics case. Although the proposed algorithms are generic in nature, we focus here on radiation transport models used in support of the design and analysis of nuclear reactor cores. In particular, we focus on replacing the traditional assembly calculations by ROM models to facilitate the generation of homogenized cross-sections for downstream core calculations. The implication is that assembly calculations could be done instantaneously therefore precluding the need for the expensive evaluation of the few-group cross-sections for all possible core conditions. Given the generic natures of the algorithms, we make an effort to introduce the material in a general form to allow non-nuclear engineers to benefit from this work.

  18. Self-Consciousness, Evaluation of Physical Characteristics, and Physical Attractiveness.

    Science.gov (United States)

    Turner, Robert G.; Gilliland, LuNell

    1981-01-01

    Investigated the relationship between public self-consciousness and speed of processing information about self. Results indicated that high public self-conciousness subjects required less time to report evaluations of their physical features. In a second study high public self-conciousness was shown to be positively related to judged physical…

  19. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  20. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  1. The Assessment Of High Temperature Reactor Fuel (Characteristics Of HTTR Fuel)

    International Nuclear Information System (INIS)

    Dewita, Erlan; Tuka, Veronica; Gunandjar

    1996-01-01

    HTTR is one of the reactor type with Helium coolant and outlet coolant temperature of 950 o C. One possibility of HTTR application is the coo generation of steam in high temperature and electric power for supply energy to industry in the future. Considering to the high operating temperature of HTTR, therefore it is needed the reactor fuel which have good mechanical, chemical and physical stability to the high temperature, and stable to the influence of fission fragment and neutron during irradiation. This assessment of the HTTR fuel characteristic based on the experiment data to find information of HTTR operation feasibility. Result of the assessment indicated that fission gas release at burn-up of 3.6 % FIMA which was the same as the maximum burn up in the HTTR design was fairly lower than the maximum release estimated in the design (5 x 10 - 4), which is R/B from the fuel fabricated by the prismatic block fuel method would be low (between 10 - 9 dan 10 - 8)

  2. Effect of Drawer Master Modeling of ZPPR15 Phase A Reactor Physics Experiment on Integral Parameter

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Kim, Sang Ji

    2011-01-01

    As a part of an International-Nuclear Engineering Research Initiative (I-NERI) Project, KAERI and ANL are analyzing the ZPPR-15 reactor physics experiments. The ZPPR-15 experiments were carried out in support of the Integral Fast Reactor (IFR) project. Because of lack of the experimental data, verifying and validating the core neutronics analysis code for metal fueled sodium cooled fast reactors (SFR) has been one of the big concerns. KAERI is developing the metal fuel loaded SFR and plans to construct the demonstration SFR by around 2028. Database built through this project and its result of analysis will play an important role in validating the SFR neutronics characteristics. As the first year work of I-NERI project, KAERI analyzed ZPPR-15 Phase A experiment among four phases (Phase A to D). The effect of a drawer master modeling on the integral parameter was investigated. The approximated benchmark configurations for each loading were constructed to be used for validating a deterministic code

  3. 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.)

  4. 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

  5. Analysis of physical characteristics of vegetable oils

    Energy Technology Data Exchange (ETDEWEB)

    Piamba Tulcan, Oscar Edwin [Universidade Nacional da Colombia (UNAL), Bogota (Colombia). Fac. de Ingenieria; Universidade Federal Fluminense (PGMEC/UFF), Niteroi, RJ (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica], E-mail: oepiambat@unal.edu.co; Andrade, Danielle Oliveira de; Andrade, Ednilton Tavares de [Universidade Federal Fluminense (TER/UFF), Niteroi, RJ (Brazil). Dept. de Engenharia Agricola e do Meio Ambiente; Pereira, Roberto Guimaraes [Universidade Federal Fluminense (TER/UFF), Niteroi, RJ (Brazil). Dept. de Engenharia Mecanica

    2008-07-01

    Different vegetable oils were characterized using standardized methods. The evaluated characteristics were density, viscosity, flow point, cloud point and corrosion. The obtained data was tabulated and compared with average composition values of oils in percentage of fatty acids and iodine number for each oil. In this analysis it is shown that viscosity decreases with the increase of the iodine number, and density decrease. The cloud and flow point have greater relation with the presence of saturated or highly unsaturated fatty acids, respectively. The index of corrosion is greater when oil saturation or its iodine number are increased. (author)

  6. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F; Leira Rey, G

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  7. Health physics aspects of advanced reactor licensing reviews

    International Nuclear Information System (INIS)

    Hinson, C.S.

    1995-01-01

    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-generationclose 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-generationclose 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-generationclose 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-generationclose quotes reactor designs currently being reviewed by the NRC

  8. 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.

  9. Study of heat exchange characteristics of the Dalat Nuclear Reactor

    International Nuclear Information System (INIS)

    An, N.K.; Huy, N.Q.

    1989-01-01

    This report is presented some experimental data and related theoretical computations concerning the thermal exchange system under normal operating or accidental conditions from the thermodynamic point of view. In the normal operation, the reactor operates under safety condition T max fuel=96.2 degree C. Under LOFA condition, the heat exchage process is still realized, therefore, we should determine the allowable limits of the thermal regime at power and at shut down condition

  10. Neutron radiation characteristics of the IVth generation reactor spent fuel

    Science.gov (United States)

    Bedenko, Sergey; Shamanin, Igor; Grachev, Victor; Knyshev, Vladimir; Ukrainets, Olesya; Zorkin, Andrey

    2018-03-01

    Exploitation of nuclear power plants as well as construction of new generation reactors lead to great accumulation of spent fuel in interim storage facilities at nuclear power plants, and in spent fuel «wet» and «dry» long-term storages. Consequently, handling the fuel needs more attention. The paper is focused on the creation of an efficient computational model used for developing the procedures and regulations of spent nuclear fuel handling in nuclear fuel cycle of the new generation reactor. A Thorium High-temperature Gas-Cooled Reactor Unit (HGTRU, Russia) was used as an object for numerical research. Fuel isotopic composition of HGTRU was calculated using the verified code of the MCU-5 program. The analysis of alpha emitters and neutron radiation sources was made. The neutron yield resulting from (α,n)-reactions and at spontaneous fission was calculated. In this work it has been shown that contribution of (α,n)-neutrons is insignificant in case of such (Th,Pu)-fuel composition and HGTRU operation mode, and integral neutron yield can be approximated by the Watt spectral function. Spectral and standardized neutron distributions were achieved by approximation of the list of high-precision nuclear data. The distribution functions were prepared in group and continuous form for further use in calculations according to MNCP, MCU, and SCALE.

  11. Heavy water reactors physics; Physique des reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Girard, Y; Lourme, P; Naudet, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors) [French] Un important programme d'etudes sur la physique des reacteurs a eau lourde est mene en France depuis assez longtemps. La decision de construire le prototype EL 4 et de s'engager ainsi dans la filiere des reacteurs a eau lourde refroidis par gaz a redonne un nouvel interet a ce programme et l'a en meme temps oriente dans une direction plus particuliere. La presente communication, rassemble les resultats des etudes faites dans ce domaine depuis la derniere conference de Geneve. Dans la premiere partie on decrit les etudes experimentales dont la plupart ont ete effectuees dans la pile d'experiences critiques Aquilon II. Les experiences sont groupees en quatre ensembles: etude systematique de reseaux (mesures de laplaciens) etudes

  12. Variation in the chemical composition, physical characteristics and ...

    African Journals Online (AJOL)

    Variation in the chemical composition, physical characteristics and energy values of cereal grains produced in the Western Cape area of South Africa. TS Brand, CW Cruywagen, DA Brandt, M Viljoen, WW Burger ...

  13. Pervious concrete physical characteristics and effectiveness in stormwater pollution reduction.

    Science.gov (United States)

    2016-04-01

    The objective of this research was to investigate the physical/chemical and water flow characteristics of various previous concrete : mixes made of different concrete materials and their effectiveness in attenuating water pollution. Four pervious con...

  14. Comparative study of soil physical characteristics of Jaipur district ...

    African Journals Online (AJOL)

    Vikram

    The present study was carried in Jaipur district of Rajasthan state to measure physical characteristics of the soil samples from different districts of Jaipur. Soils samples were taken at ..... Random field models in earth sciences. Academic. Press.

  15. Physics characteristics of CANDU cores with advanced fuel cycles

    International Nuclear Information System (INIS)

    Garvey, P.M.

    1985-01-01

    The current generation of CANDU reactors, of which some 20 GWE are either in operations or under construction worldwide, have been designed specifically for the natural uranium fuel cycle. The CANDU concept, due to its D 2 O coolant and moderator, on-power refuelling and low absorption structural materials, makes the most effective utilization of mined uranium of all currently commercialized reactors. An economic fuel cycle cost is also achieved through the use of natural uranium and a simple fuel bundle design. Total unit energy costs are achieved that allow this reactor concept to effectively compete with other reactor types and other forms of energy production. There are, however, other fuel cycles that could be introduced into this reactor type. These include the slightly enriched uranium fuel cycle, fuel cycles in which plutonium is recycled with uranium, and the thorium cycle in which U-233 is recycled. There is also a special range of fuel cycles that could utilize the spent fuel from LWR's. Two specific variants are a fuel cycle that only utilizes the spent uranium, and a fuel cycle in which both the uranium and plutonium are recycled into a CANDU. For the main part these fuel cycles are characterized by a higher initial enrichment, and hence discharge burnup, than the natural uranium cycle. For these fuel cycles the main design features of both the reactor and fuel bundle would be retained. Recently a detailed study of the use in a CANDU of mixed plutonium and uranium oxide fuel from an LWR has been undertaken by AECL. This study illustrates many of the generic technical issues associated with the use of Advanced Fuel Cycles. This paper will report the main findings of this evaluation, including the power distribution in the reactor and fuel bundle, the choice of fuel management scheme, and the impact on the control and safety characteristics of the reactor. These studies have not identified any aspects that significantly impact upon the introduction of

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

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

    2016-01-01

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) 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.

  17. 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.

  18. Research on the reactor physics using the Kyoto University Critical Assembly (KUCA)

    International Nuclear Information System (INIS)

    1986-10-01

    The Kyoto University Critical Assembly [KUCA] is a multi-core type critical assembly established in 1974, as a facility for the joint use study by researchers of all universities in Japan. Thereafter, many reactor physics experiments have been carried out using three cores (A-, B-, and C-cores) in the KUCA. In the A- and B-cores, solid moderator such as polyethylene or graphite is used, whereas light-water is utilized as moderator in the C-core. The A-core has been employed mainly in connection with the Cockcroft-Walton type accelerator installed in the KUCA, to measure (1) the subcriticality by the pulsed neutron technique for the critical safety research and (2) the neutron spectrum by the time-of-flight technique. Recently, a basic study on the tight lattice core has also launched using the A-core. The B-core has been employed for the research on the thorium fuel cycle ever since. The C-core has been employed (1) for the basic studies on the nuclear characteristics of light-water moderated high-flux research reactors, including coupled-cores, and (2) for a research related to reducing enrichment of uranium fuel used in research reactors. The C-core is being utilized in the reactor laboratory course experiment for students of ten universities in Japan. The data base of the KUCA critical experiments is generated so far on the basis of approximately 350 experimental reports accumulated in the KUCA. Besides, the assessed KUCA code system has been established through analyses on the various KUCA experiments. In addition to the KUCA itself, both of them are provided for the joint use study by researchers of all universities in Japan. (author)

  19. Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

    International Nuclear Information System (INIS)

    2002-06-01

    All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

  20. Physical characteristics of welding arc ignition process

    Science.gov (United States)

    Shi, Linan; Song, Yonglun; Xiao, Tianjiao; Ran, Guowei

    2012-07-01

    The existing research of welding arc mainly focuses on the stable combustion state and the research on the mechanism of welding arc ignition process is quite lack. The tungsten inert gas(TIG) touch arc ignition process is observed via a high speed camera and the high time resolution spectral diagnosis system. The changing phenomenon of main ionized element provided the electrons in the arc ignition is found. The metallic element is the main contributor to provide the electrons at the beginning of the discharging, and then the excitated shielding gas element replaces the function of the metallic element. The electron density during the period of the arc ignition is calculated by the Stark-broadened lines of Hα. Through the discussion with the repeatability in relaxation phenomenon, the statistical regularity in the arc ignition process is analyzed. The similar rules as above are observed through the comparison with the laser-assisted arc ignition experiments and the metal inert gas(MIG) arc ignition experiments. This research is helpful to further understanding on the generation mechanism of welding arc ignition and also has a certain academic and practical significance on enriching the welding physical theoretical foundation and improving the precise monitoring on automatic arc welding process.

  1. 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

  2. Selected physical characteristics of medical students

    Directory of Open Access Journals (Sweden)

    Dr. Lajos Ángyán

    2003-01-01

    Full Text Available The purpose of this study was to measure selected anthropometrical characteris-tics, motor abilities and cardiorespiratory functions of medical students. Eighty-seven students were involved in this investigation. The students were categorized into five groups: (1 recreational, doing sport activities irregularly, (2 basketball and (3 handball players, having training at least two times per week, as well as men (4 and women (5 students entering medical school. In all groups the mean body mass index and waist-to-hip ratio were at the upper level of the normal range, while body fat percentage was similar to standards for sedentary subjects. Better motor per-formances were obtained from the basketball and handball players than from the other groups. Static strength for the sample was somewhat above the normal sedentary level. The resting blood pressure and heart rate for most subjects were in the normal. Cardiovascular risk factors were found in six students. Their systolic blood pressure was above 140 mm Hg. There were no sub-jects identified with low blood pressure. The heart rate was elevated for three students from the recreational group, and in the women. Bradycardia did not occur. The vital capacity and the ability to hold one’s breath was at the upper level of the normal range. The present results emphasis the need to improve the students` prevention oriented life style through participation in exercising.

  3. 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.

  4. Berkeley Nuclear Laboratories Reactor Physics Mk. III Experimental Programme. Description of facility and programme for 1971

    Energy Technology Data Exchange (ETDEWEB)

    Nunn, R M; Waterson, R H; Young, J D

    1971-01-15

    Reactor physics experiments have been carried out at Berkeley Nuclear Laboratories during the past few years in support of the Civil Advanced Gas-Cooled Reactors (Mk. II) the Generating Board is building. These experiments are part of an overall programme whose objective is to assess the accuracy of the calculational methods used in the design and operation of these reactors. This report provides a description of the facility for the Mk. III experimental programme and the planned programme for 1971.

  5. Quantifying physical characteristics of wildland fuels using the fuel characteristic classification system.

    Science.gov (United States)

    Cynthia L. Riccardi; Susan J. Prichard; David V. Sandberg; Roger D. Ottmar

    2007-01-01

    Wildland fuel characteristics are used in many applications of operational fire predictions and to understand fire effects and behaviour. Even so, there is a shortage of information on basic fuel properties and the physical characteristics of wildland fuels. The Fuel Characteristic Classification System (FCCS) builds and catalogues fuelbed descriptions based on...

  6. Actual and preferred personality characteristics of physical educators

    Directory of Open Access Journals (Sweden)

    Saša Cecić Erpič

    2002-12-01

    Full Text Available The Five-Factor Model of personality, which includes dimensions energy, agreeableness, conscientiousness, emotional stability, and openness, gained a growing acceptance by personality researchers. In the present study the actual personality characteristics of physical educators and the personality profile of an ideal (according to subjective evaluations of experts physical educator were investigated. The aim of the study was to examine differences between profiles of actual and preferred personality characteristics and to present differences in personality characteristics between male and female physical educators of different ages. The study included 76 (40 male, 36 female 24 to 58 year-old physical educators (mean age 39.7 years. 34 experts from the field of sport, physical education, and kinesiology evaluated the preferred personality structure of an ideal physical educator. The Big Five Observer (BFO was used to assess actual and preferred personality structures. These results show that the actual personality profile of physical educators is described with all five moderately high dimensions, which is relatively congruent with the social type from Holland's typology, characteristic of educators. In comparison with participants, an ideal physical educator should have equally expressed agreeableness, while other four dimensions should be highly expressed. Gender differences in energy and agreeableness (women have higher results in both dimensions were found significant. No significant age differences in the personality structure were obtained by a cross-sectional comparison.

  7. ERP differences between processing of physical characteristics and personality attributes

    Science.gov (United States)

    2012-01-01

    Background Limited data from behavioral and brain-imaging studies indicate that personality traits and physical characteristics are processed differently by the brain. Additionally, electrophysiological results of studies comparing the processing of positive and negative words have produced mixed results. It is therefore not clear how physical and personality attributes with emotional valence (i.e., positive and negative valence) are processed. Thus, this study aimed to examine the neural activity associated with words describing personality traits and physical characteristics with positive or negative emotional valence using Event Related Potentials (ERPs). Methods A sample of 15 healthy adults (7 men, 8 women) participated in a computerized word categorization task. Participants were asked to categorize visual word stimuli as physical characteristics or personality traits, while ERPs were recorded synchronously. Results Behavioral reaction times to negative physical stimuli were shorter compared to negative personality words, however reaction times did not significantly differ for positive stimuli. Electrophysiological results showed that personality stimuli elicited larger P2 and LPC (Late Positive Component) amplitudes compared to physical stimuli, regardless of negative or positive valence. Moreover, negative as compared with positive stimuli elicited larger P2 and LPC amplitudes. Conclusion Personality and physical stimuli were processed differently regardless of positive or negative valence. These findings suggest that personality traits and physical characteristics are differentially classified and are associated with different motivational significance. PMID:22967478

  8. Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

    1986-01-01

    Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

  9. Fabrication characteristics of zircaloy tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Haydt, H.M.

    1980-11-01

    The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author) [pt

  10. Influence of core model parameters on the characteristics of neutron beams of the research reactor

    Directory of Open Access Journals (Sweden)

    N. A. Khafizova

    2013-12-01

    Full Text Available IRT MEPhI reactor is equipped with a number of facilities at horizontal experimental channels (HEC. Knowing of parameters influencing spatio-angular distribution of irradiation fields is essential for each application area. The research for neutron capture therapy (NCT facility at HEC of the reactor was made. Calculation methods have been used to estimate how the reactor core parameters influence neutron beam characteristics at the HEC output. The impact of neutron source model in Monte Carlo calculations by MCNP code on the parameters of neutron and secondary photon field at the output of irradiation beam tubes of research reactor is estimated. The study shows that specifying neutron source with fission reaction rate distribution in SDEF option gives almost the same results as criticality calculation considered the most accurate. Our calculations show that changes of the core operational parameters have insignificant influence on characteristics of neutron beams at HEC output.

  11. Physics analysis of the TIBER-II engineering test reactor

    International Nuclear Information System (INIS)

    Uckan, N.A.; Houlberg, W.A.; Attenberger, S.E.; Dory, R.A.; Spong, D.A.; Tolliver, J.S.; Sheffield, J.

    1987-11-01

    Confinement capability, burn characteristics, heating and fueling requirements, and fast alpha particle effects are assessed for the TIBER-II engineering test reactor (ETR/ITER). Confinement predictions for a wide variety of empirical scaling laws show that ignition in TIBER-II (or similar ETR-like devices) is marginal at 10 MA, whereas the design goal to achieve noninductively driven, steady-state burn with Q > 5 can easily be attained. Operation at the higher plasma currents being discussed for ITER or the attainment of higher density limits and/or favorable H-mode scalings improves the ignition capability. Pellet penetration calculations indicate that density profile control with pellets may not be feasible even for pellet velocities up to about 50 km/s, however, density peaking could result from inward pinch effects, as frequently inferred from experiments. The fast alpha contribution to pressure is substantial (10 to 30%) at TIBER (or any ETR/ITER) burn temperatures (8 to 20 keV). A relatively low level of fast alpha radial diffusion or a modest level of thermal alpha buildup significantly influences the ignition and steady-state burn capability. The fast alpha population can also modify the background plasma ballooning mode stability boundaries, lowering the beta limit β/sub crit/ - in particular, operation at the high electron temperatures needed for efficient current drive can exacerbate this effect. The use of high-energy neutral beams offers the promise of two important improvements in projected performance: an effective method for noninductive current drive and a means for controlling the current density profile deep within the plasma, as required for stable operation at high beta levels. 14 refs., 10 figs., 1 tab

  12. Physics analysis of the TIBER-II engineering test reactor

    International Nuclear Information System (INIS)

    Uckan, N.A.; Houlberg, W.A.; Attenberger, S.E.; Dory, R.A.; Spong, D.A.; Tolliver, J.S.; Sheffield, J.

    1987-01-01

    Confinement capability, burn characteristics, heating and fueling requirements, and fast-alpha particle effects are assessed for the TIBER-II engineering test reactor (ETR/ITER). Confinement predictions for a wide variety of empirical scaling laws show that ignition on TIBER-II (or similar ETR-like devices) is marginal at 10 MA, whereas the design goal to achieve noninductively driven, steady-state burn with Q > 5 can easily be attained. Operation at the higher plasma currents being discussed for ITER or the attainment of higher density limits and/or favorable H-mode scalings improves the ignition capability. Pellet penetration calculations indicate that density profile control with pellets may not be feasible even for pellet velocities up to bout 50 km/s; however, density peaking could result from inward pinch effects, as frequently inferred from experiments. The fast alpha contribution to pressure is substantial (10-30%) at TIBER (or any ETR/ITER) burn temperatures (8-20 keV). A relatively low level of fast alpha radial diffusion or a modest level of thermal alpha buildup significantly influences the ignition and steady-state burn capability. The fast alpha population can also modify the background plasma ballooning mode stability boundaries, lowering the beta limit β/sub crit/ - in particular, operation at the high electron temperatures needed for efficient current drive can exacerbate this effect. The use of high-energy neutral beams offers the promise of two important improvements in projected performance: an effective method for noninductive current drive and a means for controlling the current density profile deep within the plasma, as required for stable operation at high beta levels

  13. 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

  14. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  15. 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...

  16. 78 FR 69139 - Physical Security-Design Certification and Operating Reactors

    Science.gov (United States)

    2013-11-18

    ... Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan--draft section..., ``Physical Security--Design Certification and Operating Reactors.'' The public comment period was originally....regulations.gov and search for Docket ID NRC-2013-0225. Address questions about NRC dockets to Carol Gallagher...

  17. Fast neutron reactor noise analysis: beginning failure detection and physical parameter estimation

    International Nuclear Information System (INIS)

    Le Guillou, G.

    1975-01-01

    The analysis of the signals fluctuations coming from a power nuclear reactor (a breeder), by correlation methods and spectral analysis has two principal applications: on line estimation of physical parameters (reactivity coefficients); beginning failures (little boiling, abnormal mechanic vibrations). These two applications give important informations to the reactor core control and permit a good diagnosis [fr

  18. 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

  19. Evaluation of the influence of seismic restraint characteristics on breeder reactor piping systems

    International Nuclear Information System (INIS)

    Mello, R.M.; Pollono, L.P.

    1979-01-01

    For the Clinch River Breeder Reactor Plant (CRBRP) heat transport system piping within the reactor containment building, dynamic analyses of the piping loops have been performed to study the effect of restraint stiffness on the dynamic behavior of the piping. In addition, analysis and testing of typical CRBRP restraint system components have been performed for the purpose of quantifying and verifying the basic characteristics of the restraints used in the piping system dynamic analysis

  20. PHYSICAL AND MECHANICAL CHARACTERISTICS OF BUILDING MATERIALS OF HISTORIC BUILDINGS

    Directory of Open Access Journals (Sweden)

    Jiří Witzany

    2017-12-01

    Full Text Available The article presents partial results of laboratory research into physical and mechanical characteristics of materials most commonly used as walling units in masonry structures of historic and heritage buildings. Core boreholes and specimens for the laboratory research of selected characteristics were sampled from accessible places of historic buildings, which had not been restored or reconstructed. The results of the research brought new knowledge about the unreliability (variance of the properties of historical, mainly natural building materials, and, at the same time, pointed out the need for further research and extension of knowledge necessary for the assessment of residual physical and mechanical characteristics of historic masonry structures.

  1. 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

  2. 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...

  3. 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

  4. 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)

  5. Summary record of the 33. Meeting of NEA committee on reactor physics

    International Nuclear Information System (INIS)

    Martinelli, R.

    1991-01-01

    This paper is the summary record of the thirty-third meeting (Technical session) of the Nuclear Energy Agency Committee on Reactor Physics. A complete list of all the papers presented at this meeting is given in annex 4

  6. 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)

  7. 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

  8. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    International Nuclear Information System (INIS)

    Hu Jian; Jiang Nan; Li Jie; Shang Kefeng; Lu Na; Wu Yan; Mizuno Akira

    2016-01-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. (paper)

  9. Discharge Characteristics of Series Surface/Packed-Bed Discharge Reactor Diven by Bipolar Pulsed Power

    Science.gov (United States)

    Hu, Jian; Jiang, Nan; Li, Jie; Shang, Kefeng; Lu, Na; Wu, Yan; Mizuno, Akira

    2016-03-01

    The discharge characteristics of the series surface/packed-bed discharge (SSPBD) reactor driven by bipolar pulse power were systemically investigated in this study. In order to evaluate the advantages of the SSPBD reactor, it was compared with traditional surface discharge (SD) reactor and packed-bed discharge (PBD) reactor in terms of the discharge voltage, discharge current, and ozone formation. The SSPBD reactor exhibited a faster rising time and lower tail voltage than the SD and PBD reactors. The distribution of the active species generated in different discharge regions of the SSPBD reactor was analyzed by optical emission spectra and ozone analysis. It was found that the packed-bed discharge region (3.5 mg/L), rather than the surface discharge region (1.3 mg/L) in the SSPBD reactor played a more important role in ozone generation. The optical emission spectroscopy analysis indicated that more intense peaks of the active species (e.g. N2 and OI) in the optical emission spectra were observed in the packed-bed region. supported by National Natural Science Foundation of China (No. 51177007), the Joint Funds of National Natural Science Foundation of China (No. U1462105), and Dalian University of Technology Fundamental Research Fund of China (No. DUT15RC(3)030)

  10. Health physics aspects of activation products from fusion reactors

    International Nuclear Information System (INIS)

    Shoup, R.L.; Poston, J.W.; Easterly, C.E.; Jacobs, D.G.

    1975-01-01

    A review of the activation products from fusion reactors and their attendant impacts is discussed. This includes a discussion on their production, expected inventories, and the status of metabolic data on these products

  11. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    Accelerator driven systems (ADS) are attracting worldwide attention .... The region of interest (or the entire reactor core) is divided into a suitable number ..... have also presented the status of the theoretical and experimental activities being.

  12. Analysis for RSG-GAS operational characteristics of reactor cooling system

    International Nuclear Information System (INIS)

    Nurhappy, T.

    1998-01-01

    Analysis of operational characteristics of reactor cooling systems (JE01 and PA) is aimed at determining the effects of operation and maintenance patterns to the operational characteristic of the system. Analysis is carried out by virtue of the operating and maintenance data from 1987 to 1997, comprising the operating hours (duration) and data on operating failures of the systems. Results of study show that, either separately or jointly, the operating and maintenance patterns will qualitatively affect the operational characteristic of the systems

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. Development of a compact digital reactivity meter and a reactor physics data processor

    International Nuclear Information System (INIS)

    Shimazu, Y.; Nakano, Y.; Tahara, Y.; Okayama, T.

    1987-01-01

    Reactor physics tests at initial startup and after refuelings are performed to verify the nuclear design and to assure safe operation. Analog computers and instruments are widely used for the acquisition of data, and these data are reduced by hand. These conventional procedures, however, require much time and labor. Since there has been great progress in the development of digital computers and devices, these procedures are digitalized, which successfully reduces the time and labor required for reactor physics tests

  18. Analysis on flow characteristic of nuclear heating reactor

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin

    1997-06-01

    The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam mass, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction, mass flow rate and stability of the system, especially at lower pressure, secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit has not reaches its saturation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0 K<ΔT<28 K) there exists three regions for system flow condition, namely, (1) stable two-phase flow, (2) bulk and subcooled boiling unstable flow, (3) subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, is showed in the above inlet subcooling range, and based on it the instability map of the system is given through experiment and calculation. (3 refs., 9 figs.)

  19. Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

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

    International Nuclear Information System (INIS)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K.; Kim, J. H.

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  2. 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

  3. 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

  4. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

    1986-01-01

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

  5. Status of computer codes available in AEOI for reactor physics analysis

    International Nuclear Information System (INIS)

    Karbassiafshar, M.

    1986-01-01

    Many of the nuclear computer codes available in Atomic Energy Organization of Iran AEOI can be used for physics analysis of an operating reactor or design purposes. Grasp of the various methods involved and practical experience with these codes would be the starting point for interesting design studies or analysis of operating conditions of presently existing and future reactors. A review of the objectives and flowchart of commonly practiced procedures in reactor physics analysis of LWRs and related computer codes was made, extrapolating to the nationally and internationally available resources. Finally, effective utilization of the existing facilities is discussed and called upon

  6. 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

  7. 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.

  8. 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

  9. 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. (author)

  10. Characteristics and uses of a 250 kW TRIGA reactor

    International Nuclear Information System (INIS)

    Dimic, V.

    1985-01-01

    The 250 kW TRIGA Mark II reactor is a light water reactor with solid fuel elements in which the zirconium hydride moderator is homogeneously distributed between enriched uranium. Therefore the reactor has the large prompt negative temperature coefficient of reactivity, the fuel also has very high retention of radioactive fission products. The reactor core is a cylindrical configuration with an annular graphite reflector. The experimental facilities include a rotary specimen rack, a central incore radiation thimble, a pneumatic transfer system, and pulsing capability. Other experimental facilities include two radial and two tangential beam tubes, a graphite thermal column, and a graphite thermalizing column. At the steady state power of 250 kW the peak flux is 1x10 13 n/cm 2 s in the central test position. In addition, pulsing to about 2000 MW is usually provided giving peak fluxes of about 2x10 16 n/cm 2 sec. All TRIGA reactors produce a core-average thermal neutron flux of about 10 7 n.v per watt. Only with very large accelerators could such a high neutron flux be achieved. In order to give an appreciation for the research conducted at research reactors, the types of research could be summarized as follows: thermal neutron scattering, neutron radiography, neutron and nuclear physics, activation analysis, radiochemistry, biology and medicine, and teaching and training. Typical applied research with a 250 kW reactor has been conducted in medicine in biology, archeology, metallurgy and materials science, engineering and criminology. It is well known that research reactors have been used routinely to produce isotopes for industry and medicine. In some instances, reactors are the preferred method of isotope production. We can conclude that the 250 kW TRIGA research reactor is a useful and wide ranging source of radiation for basic and applied research. The operation cost for this instrument is relatively low. (author)

  11. Activity report of working party on reactor physics of accelerator-driven system. July 1999 to March 2001

    International Nuclear Information System (INIS)

    2002-02-01

    Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Accelerator-Driven System (ADS-WP) was set in July 1999 to review and investigate special subjects related to reactor physics research for the Accelerator-Driven Subcritical System (ADS). The ADS-WP, at the first meeting, discussed a guideline of its activity for two years and decided to concentrate upon three subjects: (1) neutron transport calculations in high energy range, (2) static and kinetic (safety-related) characteristics of subcritical system, and (3) system design including ADS concepts and elemental technology developments required. The activity of ADS-WP continued from July 1999 to March 2001. In this duration, the members of ADS-WP met together four times and discussed the above subjects. In addition, the ADS-WP conducted a questionnaire on requests and proposals for the plan of Transmutation Physics Experimental Facility in the High-Intensity Proton Accelerator Project, which is a joint project between JAERI and KEK (High Energy Accelerator Research Organization). This report summarizes the results obtained by the above ADS-WP activity. (author)

  12. Neutronic characteristics of linear-assembly breed-and-burn reactors

    International Nuclear Information System (INIS)

    Petroski, Robert; Forget, Benoit; Forsberg, Charles

    2012-01-01

    Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which

  13. Core management and reactor physics aspects of the conversion of the NRU reactor to LEU

    International Nuclear Information System (INIS)

    Atfield, M.D.

    1985-01-01

    Results of work done to assess the effects of converting the NRU reactor to LEU are presented. The effects are small, and the operational rules and safety analysis, appropriate to the HEU core, will still apply. (author)

  14. Study on unstable fracture characteristics of light water reactor piping

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    1998-08-01

    Many testing studies have been conducted to validate the applicability of the leak before break (LBB) concept for the light water reactor piping in the world. It is especially important among them to clarify the condition that an inside surface crack of the piping wall does not cause an unstable fracture but ends in a stable fracture propagating only in the pipe thickness direction, even if the excessive loading works to the pipe. Pipe unstable fracture tests performed in Japan Atomic Energy Research Institute had been planned under such background, and clarified the condition for the cracked pipe to cause the unstable fracture under monotonous increase loading or cyclic loading by using test pipes with the inside circumferential surface crack. This paper examines the pipe unstable fracture by dividing it into two parts. One is the static unstable fracture that breaks the pipe with the inside circumferential surface crack by increasing load monotonously. Another is the dynamic unstable fracture that breaks the pipe by the cyclic loading. (author). 79 refs

  15. Evaluation of physical and chemical characteristics of xanthan gums

    Directory of Open Access Journals (Sweden)

    Claire Tondo Vendruscolo

    2007-03-01

    Full Text Available This work aimed at evaluating the physical and chemical characteristics of the xanthan produced by Xanthomonas arboricola pv pruni strain 115, and at comparing it with two xanthans obtained from commercial sources – Kelzan e Roeper. The analyzed xanthans showed specified patterns mentioned in the literature, except for low pyruvic acid content in the xanthan produced by strain 115, low monovalent salt content in the Roeper sample and high divalent salt content in both commercial samples. The low pyruvic acid content in the xanthan produced by X. arboricola pv pruni 115 did not affect the aqueous solution viscosity. Thus, the xanthan produced by strain 115 show physical and chemical characteristics that allow its use by the petroleum industry, as well as, in food, pharmaceutical and cosmetics products. Xanthomonas arboricola pv pruni. Xanthan. Physical and chemical characteristics.

  16. Benefits of reactor physics experiments for the HTGR industrial development - an attempt to a quantitative approach

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Graziani, G; Massino, L; Rinaldini, C; Zanantoni, C

    1972-10-15

    The available results of reactor physics experiments on HTGRs and their accuracies are briefiy reviewed. The physical quantities of interest are grouped into three categories: basic nuclear data, lattice parameters and integral design data. The last two are considered and their possible improvements in accuracy by means of experimental measurements are assessed. The cost penalty on fuel cycle and capital cost due to each physical quantity is then considered, and consequently the benefits of reactor physics experiments are evaluated for a number of hypotheses concerning the foreseeable HTGR development and the delay in taking practical advantage of experimental results. It is concluded that, at the present state of knowledge of basic nuclear data and with the available calculation methods, the economic incentive to new reactor physics experiments is small, and a previous careful analysis is recommended to those intending to perform such experiments.

  17. 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.)

  18. Preliminary analysis of basic reactor physics of the Dual Fluid Reactor - 15270

    International Nuclear Information System (INIS)

    Wang, X.; Macian-Juan, R.; Seidl, M.

    2015-01-01

    The Dual Fluid Reactor (DFR) is a novel fast nuclear reactor concept invented by the IFK based on the Generation IV Molten Salt Reactor and the Liquid Metal Cooled Reactor. The DFR uses a chloride based molten fuel salt in order to harden the neutron spectrum. The molten fuel salt is cooled with a separated liquid lead loop, which in principle allows for higher power densities and better breeding performance. The DFR does not combine heat removal and breeding into a single circuit but separates the two functions into two independent circuits. Since there are attractive features mentioned in this design, the main task of this paper is to verify the model of the whole reactor based on this concept. For this purpose several calculations are presented, including steady state calculations, sensitivity calculations with regard to the nuclide cross sections, the temperature and geometry coefficient of k eff as well as the burnup calculation. The Monte Carlo calculation codes MCNP, SERPENT and SCALE are used for the analysis. As expected the study shows a significant negative reactivity feedback with temperature in the overall fission zone. For the coupled coolant and reflector design the temperature feedback is rather small for practical purposes such as reactor control during normal operation. In the view of these results the DFR in principle can be self-regulated totally by the temperature change of its own fuel salt and consequently can rely on fully passive safety systems for accident management

  19. Rotating bed reactor for CLC: Bed characteristics dependencies on internal gas mixing

    International Nuclear Information System (INIS)

    Håkonsen, Silje Fosse; Grande, Carlos A.; Blom, Richard

    2014-01-01

    Highlights: • A mathematical model for the rotating CLC reactor has been developed. • The model reflects the gas distribution in the reactor during CLC operation. • Radial dispersion in the rotating bed is the main cause for internal gas mixing. • The model can be used to optimize the reactor design and particle characteristics. - Abstract: A newly designed continuous lab-scale rotating bed reactor for chemical looping combustion using CuO/Al 2 O 3 oxygen carrier spheres and methane as fuel gives around 90% CH 4 conversion and >90% CO 2 capture efficiency based on converted methane at 800 °C. However, from a series of experiments using a broad range of operating conditions potential CO 2 purities only in the range 20–65% were yielded, mostly due to nitrogen slip from the air side of the reactor into the effluent CO 2 stream. A mathematical model was developed intending to understand the air-mixing phenomena. The model clearly reflects the gas slippage tendencies observed when varying the process conditions such as rotation frequency, gas flow and the flow if inert gas in the two sectors dividing the air and fuel side of the reactor. Based on the results, it is believed that significant improvements can be made to reduce gas mixing in future modified and scaled-up reactor versions

  20. Fluid Flow Characteristic Simulation of the Original TRIGA 2000 Reactor Design Using Computational Fluid Dynamics Code

    International Nuclear Information System (INIS)

    Fiantini, Rosalina; Umar, Efrizon

    2010-01-01

    Common energy crisis has modified the national energy policy which is in the beginning based on natural resources becoming based on technology, therefore the capability to understanding the basic and applied science is needed to supporting those policies. National energy policy which aims at new energy exploitation, such as nuclear energy is including many efforts to increase the safety reactor core condition and optimize the related aspects and the ability to build new research reactor with properly design. The previous analysis of the modification TRIGA 2000 Reactor design indicates that forced convection of the primary coolant system put on an effect to the flow characteristic in the reactor core, but relatively insignificant effect to the flow velocity in the reactor core. In this analysis, the lid of reactor core is closed. However the forced convection effect is still presented. This analysis shows the fluid flow velocity vector in the model area without exception. Result of this analysis indicates that in the original design of TRIGA 2000 reactor, there is still forced convection effects occur but less than in the modified TRIGA 2000 design.

  1. Activity report of working party on reactor physics of subcritical system. October 2001 to March 2003

    International Nuclear Information System (INIS)

    2004-03-01

    Under the Research Committee on Reactor Physics, the Working Party on Reactor Physics of Subcritical System (ADS-WP) was set in July 2001 to research reactor physics of subcritical system such as Accelerator-Driven System (ADS). The WP, at the first meeting, discussed a guideline of its activity for two years and decided to perform theoretical research for the following subjects: (1) study of reactor physics for a subcritical core, (2) benchmark problems for a subcritical core and their calculations, (3) study of physical parameters affecting to set subcriticality of ADS, and (4) study of measurement and surveillance methods of subcriticality of a subcritical core. The activity of ADS-WP continued up to March 2003. In this duration, the members of the WP met together eight times, including four meetings jointly held with the Workshop on Accelerator-Driven Subcritical Reactor at Kyoto University Research Reactor Institute. This report summarizes the result obtained by the above WP activity and research. (author)

  2. Operational characteristics of the CALIBAN fast pulse reactor

    International Nuclear Information System (INIS)

    Cortella, J.; Reberol, R.; Vanel, M.

    1976-01-01

    CALIBAN is a FPR operated at CEA-Valduc Center since 1971. It has been designed as a fast neutron irradiation source and its environment is specific for this utilization. To date, it delivered more than 400 bursts and the fuel integrated about 5.10 19 fissions. The main characteristics are: - cylindirical core 113 kg U 235 - Mo 10% alloy - integrated dose in a burst in the central 2.5cm diam cavity:3.10 14 n.cm -2 - integrated dose in a burst outside of the core:5.10 13 n.cm -2 - pulse width:50μs A special effort was made in measuring the spectrometric and dosimetric neutron and gamma characteristics. Some results will be presented here. (auth.)

  3. Physical and decay characteristics of commercial LWR spent fuel

    International Nuclear Information System (INIS)

    Roddy, J.W.; Claiborne, H.C.; Ashline, R.C.; Johnson, P.J.; Rhyne, B.T.

    1986-01-01

    Information was collected from the literature and from major manufacturers that will be useful in the design and construction of a mined geologic repository for the disposal of light-water-reactor spent fuel. Pertinent data are included on mechanical design characteristics and materials of construction for fuel assemblies and fuel rods and computed values for heat generation rates, radioactivity, and photon and neutron emission rates as a function of time for four reference cases. Calculations were made with the ORIGEN2 computer code for burnups of 27,500 and 40,000 MWd for a typical boiling-water reactor and 33,000 and 60,000 MWd for a typical pressurized-water reactor. The results are presented in figures depicting the individual contributions per metric ton of initial heavy metal for the activation products, fission products, and actinides and their daughters to the radioactivity and thermal power as a function of time. Tables are also presented that list the contribution of each major nuclide to the radioactivity, thermal power, and photons and neutrons emitted for disposal emitted for disposal periods from 1 to 100,000 years

  4. Uncertainties in HTGR neutron-physical characteristics due to computational errors and technological tolerances

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Grebennik, V.N.; Davidenko, V.G.; Kosovskij, V.G.; Smirnov, O.N.; Tsibul'skij, V.F.

    1991-01-01

    The paper is dedicated to the consideration of uncertainties is neutron-physical characteristics (NPC) of high-temperature gas-cooled reactors (HTGR) with a core as spherical fuel element bed, which are caused by calculations from HTGR parameters mean values affecting NPC. Among NPC are: effective multiplication factor, burnup depth, reactivity effect, control element worth, distribution of neutrons and heat release over a reactor core, etc. The short description of calculated methods and codes used for HTGR calculations in the USSR is given and evaluations of NPC uncertainties of the methodical character are presented. Besides, the analysis of the effect technological deviations in parameters of reactor main elements such as uranium amount in the spherical fuel element, number of neutron-absorbing impurities in the reactor core and reflector, etc, upon the NPC is carried out. Results of some experimental studies of NPC of critical assemblies with graphite moderator are given as applied to HTGR. The comparison of calculations results and experiments on critical assemblies has made it possible to evaluate uncertainties of calculated description of HTGR NPC. (author). 8 refs, 8 figs, 6 tabs

  5. Applications of Oregon State University's TRIGA reactor in health physics education

    International Nuclear Information System (INIS)

    Higginbotham, J.F.

    1990-01-01

    The Oregon State University TRIGA reactor (OSTR) is used to support a broad range of traditional academic disciplines, including anthropology, oceanography, geology, physics, nuclear chemistry, and nuclear engineering. However, it also finds extensive application in the somewhat more unique area of health physics education and research. This paper summarizes these health physics applications and briefly describes how the OSTR makes important educational contributions to the field of health physics

  6. Characteristics of Missing Physical Activity Data in Children and Youth

    Science.gov (United States)

    Zhuang, Jie; Chen, Peijie; Wang, Chao; Huang, Liang; Zhu, Zheng; Zhang, Wenjie; Fan, Xiang

    2013-01-01

    Purpose: The purpose of this study was to investigate the characteristics of missing physical activity (PA) data of children and youth. Method: PA data from the Chinese City Children and Youth Physical Activity Study ("N" = 2,758; 1,438 boys and 1,320 girls; aged 9-17 years old) were used for the study. After the data were sorted by the…

  7. Structural Characteristics and Physical Properties of Tectonically Deformed Coals

    OpenAIRE

    Yiwen Ju; Zhifeng Yan; Xiaoshi Li; Quanlin Hou; Wenjing Zhang; Lizhi Fang; Liye Yu; Mingming Wei

    2012-01-01

    Different mechanisms of deformation could make different influence on inner structure and physical properties of tectonically deformed coal (TDC) reservoirs. This paper discusses the relationship between macromolecular structure and physical properties of the Huaibei-Huainan coal mine areas in southern North China. The macromolecular structure and pore characteristics are systematically investigated by using techniques such as X-ray diffraction (XRD), high-resolution transmission electron mic...

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

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

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

  9. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

  10. 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

  11. Complementarity of integral and differential experiments for reactor physics purposes

    International Nuclear Information System (INIS)

    Tellier, Henry.

    1981-04-01

    In this paper, the following topics are studied: uranium 238 effective integral; thermal range uranium 238 capture cross section; Americium 242 m capture cross section. The mentioned examples show that differential and integral experiments are both useful to the reactor physicists

  12. Parametric study on thermal-hydraulic characteristics of high conversion light water reactor

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Fujii, Sadao.

    1988-11-01

    To assess the feasibility of high conversion light water reactors (HCLWRs) from the thermal-hydraulic viewpoint, parametric study on thermal-hydraulic characteristics of HCLWR has been carried out by using a unit cell model. It is assumed that a HCLWR core is contained in a current 1000 MWe PWR plant. At the present study, reactor core parameters such as fuel pin diameter, pitch, core height and linear heat rate are widely and parametrically changed to survey the relation between these parameters and the basic thermal-hydraulic characteristics, i.e. maximum fuel temperature, minimum DNBR, reduction of reactor thermal output and so on. The validity of the unit cell model used has been ensured by comparison with the result of a subchannel analysis carried out for a whole core. (author)

  13. Characteristics of D-3He fueled frc reactor: ARTEMIS-L

    International Nuclear Information System (INIS)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

  14. Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

    Science.gov (United States)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

  15. Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

    Energy Technology Data Exchange (ETDEWEB)

    Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

    1993-11-01

    The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

  16. A review of reactor physics uncertainties and validation requirements for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Baxter, A.M.; Lane, R.K.; Hettergott, E.; Lefler, W.

    1991-01-01

    The important, safety-related, physics parameters for the low-enriched Modular High-Temperature gas-Cooled Reactor (MHTGR) such as control rod worth, shutdown margins, temperature coefficients, and reactivity worths, are considered, and estimates are presented of the uncertainties in the calculated values of these parameters. The basis for the uncertainty estimate in several of the important calculated parameters is reviewed, including the available experimental data used in obtaining these estimates. Based on this review, the additional experimental data needed to complete the validation of the methods used to calculate these parameters is presented. The role of benchmark calculations in validating MHTGR reactor physics data is also considered. (author). 10 refs, 5 figs, 3 tabs

  17. 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

  18. Measurement of transient hydrodynamic characteristics of the reactor RA primary cooling system

    International Nuclear Information System (INIS)

    Jovic, L.; Majstorovic, D.; Zeljkovic, I.

    1987-01-01

    Experimental study of transient hydrodynamic characteristics of the research nuclear reactor RA by simultaneous measurements of fluid flow and pressure on several locations of the RA primary coolant system is done. Loss of electric power transient on the main circulation pumps is simulated. measurement methodology, data processing and results of measured data analysis are given. (author)

  19. Fast Reactor Physics. Vol. II. 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)

  20. 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.

  1. An overview of the current status of resonance theory in reactor physics applications

    International Nuclear Information System (INIS)

    Hwang, R.N.

    1993-01-01

    The neutron resonance phenomena constitute one of the most fundamental subjects in nuclear physics as well as in reactor physics. It is the area where the concepts of nuclear interaction and the treatment of the neutronic balance in reactor lattices become intertwined. The later requires the detailed knowledge of resonance structures of many nuclide of practical interest to the development of nuclear energy. The key issue of the resonance treatment in reactor applications is directly associated with the use of the microscopic cross sections in the macroscopic reactor cells with a wide range of composition, temperature,and geometric configurations. It gives rise to the so called self-shielding effect. The accurate estimations of such a effect is essential not only in the calculation of the criticality of a reactor but also from the point of view of safety considerations. The latter manifests through the Doppler effect particularly crucial to the fast reactor development. The task of accurate treatment of the self-shielding effect, however, is by no means simple. In fact, it is perhaps the most complicated problem in neutron physics which, strictly speaking, requires the dependence of many physical variables. Two important elements of particular interest are : (1) a concise description of the resonance cross sections as a function of energy and temperature; (2) accurate estimation of the corresponding neutron flux where appropriate. These topics will be discussed from both the historical as well as the state-of-art perspectives

  2. 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

  3. 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.)

  4. Physical and ionic characteristics in water soluble fraction (WSF) of ...

    African Journals Online (AJOL)

    The values of ionic and physical characteristics at 25, 50 and 100% water soluble fraction (WSF) of Olomoro well-head crude oil before and after exposure to Azolla africana were investigated. The WSF values before and after exposure to the plants showed that more ions were available after the introduction of the test plant.

  5. Oak Ridge Reservation. Physical Characteristics and National Resources

    Energy Technology Data Exchange (ETDEWEB)

    Parr, Patricia Dreyer [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Joan, F. Hughes [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2006-10-09

    The topology, geology, hydrology, vegetation, and wildlife of the Oak Ridge Reservation (ORR) provide a complex and intricate array of resources that directly impact land stewardship and use decisions. The purpose of this document is to consolidate general information regarding the natural resources and physical characteristics of the ORR.

  6. 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

  7. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

    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

  9. 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

  10. 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

  11. Multi-component controllers in reactor physics optimality analysis

    International Nuclear Information System (INIS)

    Aldemir, T.

    1978-01-01

    An algorithm is developed for the optimality analysis of thermal reactor assemblies with multi-component control vectors. The neutronics of the system under consideration is assumed to be described by the two-group diffusion equations and constraints are imposed upon the state and control variables. It is shown that if the problem is such that the differential and algebraic equations describing the system can be cast into a linear form via a change of variables, the optimal control components are piecewise constant functions and the global optimal controller can be determined by investigating the properties of the influence functions. Two specific problems are solved utilizing this approach. A thermal reactor consisting of fuel, burnable poison and moderator is found to yield maximal power when the assembly consists of two poison zones and the power density is constant throughout the assembly. It is shown that certain variational relations have to be considered to maintain the activeness of the system equations as differential constraints. The problem of determining the maximum initial breeding ratio for a thermal reactor is solved by treating the fertile and fissile material absorption densities as controllers. The optimal core configurations are found to consist of three fuel zones for a bare assembly and two fuel zones for a reflected assembly. The optimum fissile material density is determined to be inversely proportional to the thermal flux

  12. Interaction Between Physical Environment, Social Environment, and Child Characteristics in Determining Physical Activity at Child Care

    NARCIS (Netherlands)

    Gubbels, J.S.; Kremers, S.P.J.; Kann, D.H.H. van; Stafleu, A.; Candel, M.J.J.M.; Dagnelie, P.C.; Thijs, C.; Vries, N.K.de

    2011-01-01

    Objective: To investigate the association between the child-care environment and physical activity of 2- and 3-year-olds. Based on an ecological view of environmental influences on health behavior, we hypothesized that the social and physical environment, as well as child characteristics (age and

  13. 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.)

  14. Anthropometric and physical fitness characteristics of young male soccer players

    Directory of Open Access Journals (Sweden)

    Leslie Andrews Portes

    2010-06-01

    Full Text Available The objective of this study was to determine anthropometric and physical fitnesscharacteristics of Brazilian male children and adolescents at the beginning of soccer training. Inthis study, 282 male soccer players ranging in age from 10 to 13 years were evaluated. The athletesparticipated in a formal soccer training program 3 times per week, with each training lasting 3hours. Anthropometric and physical fitness parameters were obtained. The boys were divided intoage classes and prevalence data were analyzed using Pearson’s chi-square test. Parametric datawere compared by one-way ANOVA or the Kruskal-Wallis test, when necessary. The resultsare expressed as the mean ± standard deviation and a p value <0.05 was considered to be significant.Growth, development, body adiposity and physical fitness characteristics were adequateand proportional to age among the boys studied (p<0.05. It was concluded that anthropometricand physical fitness characteristics of young male elite soccer players improve with and areproportional to age. Children and adolescents greatly benefit from regular physical activity. Thepresent results show that young male soccer players present adequate anthropometric conditionsand physical fitness prior to the initiation of formal training at soccer clubs.

  15. 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

  16. 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)

  17. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  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. Thermal characteristics analysis of microwaves reactor for pyrolysis of used cooking oil

    Science.gov (United States)

    Anis, Samsudin; Shahadati, Laily; Sumbodo, Wirawan; Wahyudi

    2017-03-01

    The research is objected to develop microwave reactor for pyrolysis of used cooking oil. The effect of microwave power as well as addition of char as absorber towards its thermal characteristic were investigated. Domestic microwave was modified and used to test the thermal characteristic of used cooking oil in the terms of temperature evolution, heating rate, and thermal efficiency. The samples were examined under various microwave power of 347W, 399W, 572W and 642W for 25 minutes of irradiation time. The char loading was tested in the level of 0, 50, and 100 g. Microwave reactor consists of microwave unit with a maximum power of 642W, a ceramic reactor, and a condenser equipped with temperature measurement system was successfully developed. It was found that microwave power and addition of absorber significantly influenced the thermal characteristic of microwave reactor. Under investigated condition, the optimum result was obtained at microwave power of 642W and 100 g of char. The condition was able to provide temperature of 480°C, heating rate of 18.2°C/min and thermal efficiency of 53% that is suitable to pyrolyze used cooking oil.

  20. Non-linearity consideration when analyzing reactor noise statistical characteristics. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kebadze, B V; Adamovski, L A

    1975-06-01

    Statistical characteristics of boiling water reactor noise in the vicinity of stability threshold are studied. The reactor is considered as a non-linear system affected by random perturbations. To solve a non-linear problem the principle of statistical linearization is used. It is shown that the halfwidth of resonance peak in neutron power noise spectrum density as well as the reciprocal of noise dispersion, which are used in predicting a stable operation theshold, are different from zero both within and beyond the stability boundary the determination of which was based on linear criteria.

  1. Physical principle and engineering features of the deep pool reactor for residential heating

    International Nuclear Information System (INIS)

    Shi Gong; Zhao Zhaoyi; Guo Jingren; Tian Jiafu

    1999-01-01

    The use of nuclear energy for low temperature heating is confronted with challenges of safety and economy. The deep pool reactor, a low temperature heating reactor based on novel design principles, has been studied in detail. Results show that it has excellent safety and economic features, and is very suitable for low temperature heating purposes. The whole heating system including the nuclear reactor will be a simple and easy engineering system with the characteristics of reliability, safety and economy because the system and all its devices are based on low temperature and ordinary pressure

  2. The study on the mechanical characteristics of concrete of nuclear reactor containment structure

    International Nuclear Information System (INIS)

    Jung, W. S.; Kwon, K. J.; Cho, M. S.; Song, Y. C.

    2000-01-01

    Reactor containment structure of nuclear power plant designed by prestressed concrete causes time-dependent prestress loss due to the mechanical characteristics of concrete. Prestress loss strongly affects to the safety factor of structure under the circumstances of designing, construction and inspection. Thus, this study is to investigate the mechanical characteristics of reactor containment concrete structure of Yonggwang No. 5 and 6. In this study, the compressive strength, modulus of elasticity, poisson's ratio and creep test followed by ASTM code are performed to investigate the mechanical characteristics of concrete made by V type cement. Additionally, since creep causes more time-dependent prestress loss than the other, the measurement value from the creep test is compared with the results from the creep prediction equations by KSCE, JSCE, Hansen, ACI and CEB-FIP model for the effective application. Hereafter, the results of this study may enable to assist the calculation effective stress considering time-dependent prestress loss of the prestressed concrete structures

  3. Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

  4. Graphite reactor physics; Physique des piles a graphite

    Energy Technology Data Exchange (ETDEWEB)

    Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Noc, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    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{sup 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) [French] Entreprise il y a dix ans a l'occasion de la construction des piles de Marcoule, l'etude de la

  5. Physical Characteristics Of Female Basketball Players According To Playing Position

    Directory of Open Access Journals (Sweden)

    Kucsa Richard

    2015-05-01

    Full Text Available Physical characteristics play an important role in the selection of young basketball players and the progress in their playing performance. The aim of the study was to analyze differences in chosen physical characteristics of Slovak U17 female basketball players with respect to their playing positions. We assumed, that there will be statistically significant differences between playing positions in each performance tests results. Chosen characteristics were analyzed for 14 players (mean/SD, age 16.34±0.82; body height 179.72±8.04 cm; body weight 67.62±7.10 kg; body fat 16.59±2.04 %; VO2max 46.20±4.71 ml.kg−1.min−1 according to their playing positions (guard, forward, center. Five specific performance tests for each player were conducted as a 3/4 Basketball court sprint, 10 × 5m Shuttle test, Lane agility drill, No-step vertical jump and Maximum vertical jump. The differences in tests results by playing positions were evaluated by one-way ANOVA. There were no significant differences found in results of chosen performance tests between playing positions (p>.05. In spite of fact, that there were no significant differences, we found interesting results between playing positions in physical characteristics. Guards had better results in speed (3.73±0.16 s, quickness (17.43±0.56 s and both lower-body power tests (47.16±3.06 cm; 57.00±3.40 cm than forwards and centers. Forwards had the best results in agility test (12.54±0.43 s. The results of this study produce useful information about physical characteristics of young basketball players according to their playing position and help to diagnose and improve their performance.

  6. Qualification of the Taiwan Power Company's pressurized water reactor physics methods using CASMO-4/SIMULATE-3

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Hsien-Chuan, E-mail: linsc@iner.org.tw [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yaur, Shung-Jung; Lin, Tzung-Yi; Kuo, Weng-Sheng; Shiue, Jin-Yih; Huang, Yu-Lung [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Studsvik's core management system (CMS) was applied to Taiwan Power Company's pressurized water reactor. Black-Right-Pointing-Pointer Advanced calculation model of shutdown cooling, B-10 depletion and integrated pin exposure were introduced. Black-Right-Pointing-Pointer Core characteristic parameters such as boron letdown, low power physics test (LPPT) predictions, and reaction rate were validated to measurement data. Black-Right-Pointing-Pointer The uncertainty of each item was quantified. - Abstract: This paper presents the validation of Studsvik core management system (CMS) for application to the Maanshan units 1 and 2 reactor core physics analysis (Huang and Yang, 1994). The methodology was validated by demonstrating the ability to obtain accurate and reliable results for various conditions and applications. Core characteristic parameters such as boron letdown, low power physics test (LPPT) predictions, and reaction rate were validated. Analytical results have been compared to measured data and reliability factors of the method have been quantified.

  7. Thermo-hydraulic characteristics of ship propulsion reactor in the conditions of ship motions and safety assessment

    International Nuclear Information System (INIS)

    Kobayashi, Michiyuki; Aya, Izuo; Inasaka, Fujio; Murata, Hiroyuki; Odano, Naoteru; Shiozaki, Koki

    1998-01-01

    A research project from 1995-1999 had a plan to make experimental studies on (1) safety of nuclear ship loaded with an integral ship propulsion reactor (2) effects of pulsating flow on the thermo-hydraulic characteristics of ship propulsion reactor and (3) thermo-hydraulic behaviors of the reactor container at the time of accident in a passively safe ship propulsion reactor. Development of a data base for ship propulsion reactor was attempted using previous experimental data on the thermo-hydraulic characteristics of the reactor in the institute in addition to the present results aiming to make general analytical evaluation for the safety of the engineering-simulation system for nuclear ship. A general data base was obtained by integrating the data list and the analytical program for static characteristics. A test equipment which allows to visualize the pulsating flow was produced and visualization experiments have started. (M.N.)

  8. Physical particularities of nuclear reactors using heavy moderators of neutrons

    International Nuclear Information System (INIS)

    Kulikov, G. G.; Shmelev, A. N.

    2016-01-01

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using "2"3"3U as a fissile nuclide and "2"3"2Th and "2"3"1Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  9. Physical particularities of nuclear reactors using heavy moderators of neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

    2016-12-15

    In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

  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. 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

  12. Characteristics of joint physical custody families in Flanders

    Directory of Open Access Journals (Sweden)

    An Katrien Sodermans

    2013-04-01

    Full Text Available BACKGROUND Research conducted in the 1990s showed that children who live alternately with their mother and father after divorce (joint physical custody have closer relationships with both parents and better emotional outcomes. In 1995 and 2006, joint legal custody and joint physical custody became the default judicial recommendations in Belgium. These defaults served to increase the incidence of joint custody arrangements. However, parents with joint physical custody arrangements who divorced before 2006 may have had higher socio-economic standing and lower conflict relationships than couples that divorced afterwards. Thus earlier research on the impact of joint physical custody arrangements on child outcomes may be too optimistic when considering children of recently divorced parents. OBJECTIVE This study examines the characteristics of four different physical custody arrangements (sole mother, sole father, joint physical, and flexible custody in Flanders, Belgium, and whether these characteristics have changed over time. The legal changes in 1995 and 2006 are used to distinguish three divorce cohorts. METHODS We use data on 2,207 couples that legally divorced between 1971 and 2010 from the Divorce in Flanders project, a large-scale representative multi-actor survey. Multinomial logistic regression models provide estimates of the likelihoods of different physical custody arrangements. RESULTS The incidence of sole mother custody has decreased over the last decades and children increasingly alternate between the households of the mother and the father after divorce. The incidence of sole father custody has remained low. Higher educated parents are more likely to have joint physical custody arrangements than parents from lower social classes. Also, within couples, relative educational levels are important because the higher educated spouse is more likely to have physical custody of the child. We also find that the associations between socio

  13. Research reactor RB, technical characteristics and experimental possibilities; Zbornik radova, Konferencija o koriscenju nuklearnih reaktora u Jugoslaviji

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Vranic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1978-05-15

    Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) Istrazivacki nuklearni reaktor RB u Laboratoriji za nuklearnu energetiku i tehnicku fiziku Instituta za nuklearne nauke 'Boris Kidric' u Vinci je prvi reaktorski sistem izgradjen u Jugooslaviji 1958. godine. U ovom radu opisane su osnovne tehnicke karakteristike tog reaktora, kao i mogucnosti za izvodjenje eksperimenata koje on pruza korisnicima. Njegova relativno jednostavna konstrukcija i fleksibilnost omogucavaju da se na njemu izvrse direktna merenja niza fizickih parametara, a s druge strane odsustvo bioloskog zastitnog omotaca cini ga veoma pogodnim za razna bioloska i druga ozracivanja, a takodje i dozimetrijska merenja gde se zahteva snazan izvor neutrona. (author)

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

    International Nuclear Information System (INIS)

    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.)

  15. From fundamental mode to the PWR type reactors blow off: physical analysis and contribution to the qualification of calculation tools

    International Nuclear Information System (INIS)

    Maghnouj, A.

    1996-01-01

    The work reported in this thesis centres on the resolution of reactor physics problems posed by the use in pressurised water reactors of fuel assemblies containing mixed uranium-plutonium oxide fuel (MOX). The work is essentially dependent on the results of the EPICURE experimental programme carried out between 1988 and 1994 in the reactor EOLE at the Cadarache Research Centre of the CEA. Our contribution to the validation of the computer program APOLLO2 and of its nuclear data library CEA93 shows that this code system satisfactorily calculates the neutronic characteristics of PWR cores. The validation of the experiments has provided useful information concerning the modifications required to be made to the library CEA93, which is based on the basic library of evaluated nuclear data, JEF2. This approach should now be extended to a wider basis of reactor experimental data. The studies of methods for calculating coolant voiding coefficients has made it possible to select suitable methods based on the available deterministic methods of transport theory in 2 ad 3 dimensions. These schemes have given results in satisfactory agreement with the measurements made in EPICURE programme for both local and total coolant voiding. It would now be worth while to validate the chosen methods by comparisons with calculations made using continuous energy Monte Carlo methods. (author)

  16. Characteristics and economy of the European reactor of pressurized water (EPR)

    International Nuclear Information System (INIS)

    Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C.

    2005-01-01

    The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

  17. Characteristics of self-powered neutron detectors used in power reactors

    International Nuclear Information System (INIS)

    Todt, William H. Sr.

    1998-01-01

    Self-powered neutron detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors worldwide. This paper describes the basic properties of these radiation sensors including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs, which are being effectively, used in in-core instrumentation systems for pressurized water, heavy water and graphite moderated light water reactors. Also examples are shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurized water and heavy water reactors worldwide. (author)

  18. On the determination of boiling water reactor characteristics by noise analysis

    International Nuclear Information System (INIS)

    Kleiss, J.

    1983-01-01

    In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

  19. Steam explosion - physical foundations and relation to nuclear reactor safety

    International Nuclear Information System (INIS)

    Schumann, U.

    1982-08-01

    'Steam explosion' means the sudden evaporation of a fluid by heat exchange with a hotter material. Other terms are 'vapour explosion', 'thermal explosion', and 'energetic fuel-coolant interaction (FCI)'. In such an event a large fraction of the thermal energy initially stored in the hot material may possibly be converted into mechanical work. For pressurized water reactors one discusses (e.g. in risk analysis studies) a core melt-down accident during which molten fuel comes into contact with water. In the analysis of the consequences one has to investigate steam explosions. In this report an overview over the state of the knowledge is given. The overview is based on an extensive literature review. The objective of the report is to provide the basic knowledge which is required for understanding of the most important theories on the process of steam explosions. Following topics are treated: overview on steam explosion incidents, work potential, spontaneous nucleation, concept of detonation, results of some typical experiments, hydrodynamic fragmentation of drops, bubbles and jets, coarse mixtures, film-boiling, scenario of a core melt-down accident with possible steam-explosion in a pressurized water reactor. (orig.) [de

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

    International Nuclear Information System (INIS)

    Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

    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)

  1. Study on dynamic characteristics of reduced analytical model for PWR reactor internal structures

    International Nuclear Information System (INIS)

    Yoo, Bong; Lee, Jae Han; Kim, Jong Bum; Koo, Kyeong Hoe

    1993-01-01

    The objective of this study is to establish the procedure of the reduced analytical modeling technique for the PWR reactor internal(RI) structures and to carry out the sensitivity study of the dynamic characteristics of the structures by varying the structural parameters such as the stiffness, the mass and the damping. Modeling techniques for the PWR reactor internal structures and computer programs used for the dynamic analysis of the reactor internal structures are briefly investigated. Among the many components of RI structures, the dynamic characteristics for CSB was performed. The sensitivity analysis of the dynamic characteristics for the reduced analytical model considering the variations of the stiffnesses for the lower and upper flanges of the CSB and for the RV Snubber were performed to improve the dynamic characteristics of the RI structures against the external loadings given. In order to enhance the structural design margin of the RI components, the nonlinear time history analyses were attempted for the RI reduced models to compare the structural responses between the reference model and the modified one. (Author)

  2. 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.

  3. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    In recent years, engineering oriented work, rather than basic research and development (R&D), has led to significant progress in improving the economics of innovative fast reactors and associated fuel cycle facilities, while maintaining and even enhancing the safety features of these systems. Optimization of plant size and layout, more compact designs, reduction of the amount of plant materials and the building volumes, higher operating temperatures to attain higher generating efficiencies, improvement of load factor, extended core lifetimes, high fuel burnup, etc. are good examples of achievements to date that have improved the economics of fast neutron systems. The IAEA, through its Technical Working Group on Fast Reactors (TWG-FR) and Technical Working Group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), devotes many of its initiatives to encouraging technical cooperation and promoting common research and technology development projects among Member States with fast reactor and advanced fuel cycle development programmes, with the general aim of catalysing and accelerating technology advances in these fields. In particular the theme of fast reactor deployment, scenarios and economics has been largely debated during the recent IAEA International Conference on Fast Reactors and Related Fuel Cycles: Safe Technologies and Sustainable Scenarios, held in Paris in March 2013. Several papers presented at this conference discussed the economics of fast reactors from different national and regional perspectives, including business cases, investment scenarios, funding mechanisms and design options that offer significant capital and energy production cost reductions. This Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics addresses Member States’ expressed need for information exchange in the field, with the aim of identifying the main open issues and launching possible initiatives to help and

  4. 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

  5. Physical models and numerical methods of the reactor dynamic computer program RETRAN

    International Nuclear Information System (INIS)

    Kamelander, G.; Woloch, F.; Sdouz, G.; Koinig, H.

    1984-03-01

    This report describes the physical models and the numerical methods of the reactor dynamic code RETRAN simulating reactivity transients in Light-Water-Reactors. The neutron-physical part of RETRAN bases on the two-group-diffusion equations which are solved by discretization similar to the TWIGL-method. An exponential transformation is applied and the inner iterations are accelerated by a coarse-mesh-rebalancing procedure. The thermo-hydraulic model approximates the equation of state by a built-in steam-water-table and disposes of options for the calculation of heat-conduction coefficients and heat transfer coefficients. (Author) [de

  6. 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.

  7. 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.

  8. Problems in the assessment of inherent safety characteristics of nuclear reactors

    International Nuclear Information System (INIS)

    Garribba, S.F.; Vivante, C.

    1988-01-01

    A number of proposals are being made for an increased RD and D effort on advanced nuclear power reactors that would display outstanding safety performance. A common characteristic of the different reactor concepts would be their limited reliance upon active engineered systems under major accident conditions. However, when submitted to a more close scrutiny reactor concept options may reveal diverging safety behaviors and also development opportunities. In this respect, three issues are explored in this paper. A first question is the meaning of non-active, i.e. inherent and passive safety features. Next, is the ranking of advanced and new reactor concepts from the viewpoint of inherent and passive safety. Multiple correspondence analysis may provide a simple tool, whose use is shown for the case of HTR-500, AP600 and PRISM. Conversely, probabilistic risk assessment would allow quantitative comparisons, although lack of information and data is an obstacle. Finally, is demonstration of safety performances as a step toward market deployment of the new reactor systems

  9. PHYSICAL EDUCATION AND INDIVIDUAL CHARACTERISTICS OF THE AGE SPECIFIC DEVELOPMENT

    Directory of Open Access Journals (Sweden)

    Е. М. Revenko

    2017-01-01

    Full Text Available The aim of this paper is scientific substantiation of the importance of the individual characteristics of the age specific youth development which will result in the rational modelling of students’ physical education.Methodology and research methods. The methods involve collection of experimental data carried out by means of evaluation of motor abilities and general intelligence of students. Motor abilities of students were studied by measuring of strength (dead lift dynamometry, strength endurance (pull-up, speed and power abilities (standing jump, as well as speed ability (running 30, 60 or 100 m, depending on age, aerobic endurance (running 1000 or 3000 m, depending on age. The dynamics of integral physical preparedness (DIPP of each student was calculated by calculation the arithmetic mean values of the growth rates of the development of motor abilities. Assessment of General Intelligence (GI of the 8th, 10th and 11th-grades school pupils as well as the 1st to 3rd year students was carried out through the test of R. Amthauer in the adaptation of L. A. Yazykova, and school pupils of the 6th grade were assessed through the Intelligent Test (GIT.Results. Discrepancies in the dynamics of the mental and motor areas development of maturing personality, which are interpreted as individual characteristics of the age specific development are experimentally revealed. Individual psychological differences leading to the different susceptibility to the development of motor and intellectual abilities appearing in adolescence and early adolescence are analysed. A leading role of activity in formation of the individual characteristics of the age specific development is substantiated. The conclusion of necessity to formulate to the students differing in individual characteristics of the age specific development differentiated in the complexity requirements and motor tasks in the course of physical training is made.Scientific novelty. For the first time

  10. Nuclear power plant design characteristics. Structure of nuclear power plant design characteristics in the IAEA Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    2007-03-01

    One of the IAEA's priorities has been to maintain the Power Reactor Information System (PRIS) database as a viable and useful source of information on nuclear reactors worldwide. To satisfy the needs of PRIS users as much as possible, the PRIS database has included also a set of nuclear power plant (NPP) design characteristics. Accordingly, the PRIS Technical Meeting, organized in Vienna 4-7 October 2004, initiated a thorough revision of the design data area of the PRIS database to establish the actual status of the data and make improvements. The revision first concentrated on a detailed review of the design data completion and the composition of the design characteristics. Based on the results of the review, a modified set and structure of the unit design characteristics for the PRIS database has been developed. The main objective of the development has been to cover all significant plant systems adequately and provide an even more comprehensive overview of NPP unit designs stored in the PRIS database

  11. Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

    2012-01-01

    The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

  12. 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

  13. Model tests and numerical analysis on restoring force characteristics of reactor buildings

    International Nuclear Information System (INIS)

    Uchiyama, Y.; Suzuki, S.; Akino, K.

    1987-01-01

    Seismic shear walls of nuclear reactor buildings are composed of cylindrical, truncated cone-shape, box-shape, irregular polygonal walls or its combination and they are generally heavily reinforced concrete (RC) walls. So the elasto-plastic behaviors of those RC structures in ultimate regions have many unsolved and may be considered as especially important factors for explaining nonlinear response of nuclear reactor buildings. Following these research demands, the authors have prepared a nonlinear F.E.M. code called ''SANREF'' and made an extensive study for the restoring force characteristics of the inner concrete structures (I/C) of a PWR-type containment vessel and the principal seismic shear walls of a BWR-type reactor building by some series of reduced model tests and simulation analysis for the tests results. The detailed objectives of this study can be summarized as follows: (1) Examine the effectiveness of the configurations of shear walls, reinforcement ratios, shear span ratios (M/Qd) and vertical axial stress by ''partial model test'' which simulates some independent shear walls of the PWR-type and BWR-type reactor buildings. (2) Obtain fundamental data of restoring force characteristics of the complex shaped RC structures by ''composite model test'' which models are composed of the partial model test specimens. (3) Verify the applicability of analytical methods and constitutive modelings in SANREF code for complex shaped RC structures through nonlinear simulation analysis for the composite model test

  14. Characteristics of self-powered neutron detectors used in power reactors

    International Nuclear Information System (INIS)

    Todt, W.H.

    1997-01-01

    Self-Powered Neutron Detectors have been used effectively as in-core flux monitors for over twenty-five years in nuclear power reactors world-wide. The basic properties of these radiation sensors are described including their nuclear, electrical and mechanical characteristics. Recommendations are given for the proper choice of the self-powered detector emitter to provide the proper response time and radiation sensitivity desired for use in an effective in-core radiation monitoring system. Examples are shown of specific self-powered detector designs which are being effectively used in in-core instrumentation systems for pressurised water, heavy water and graphite moderated light water reactors. Examples are also shown of the mechanical configurations of in-core assemblies of self-powered detectors combined with in-core thermocouples presently used in pressurised water and heavy water reactors worldwide. This paper is a summary of a new IEC standard to be issued in 1996 describing the characteristics and test methods of self-powered detectors used in nuclear power reactors. (author)

  15. Core characteristics on a hybrid type fast reactor system combined with proton accelerator

    International Nuclear Information System (INIS)

    Kowata, Yasuki; Otsubo, Akira

    1997-06-01

    In our study on a hybrid fast reactor system, we have investigated it from the view point of transmutation ability of trans-uranium (TRU) nuclide making the most effective use of special features (controllability, hard neutron spectrum) of the system. It is proved that a proton beam is superior in generation of neutrons compared with an electron beam. Therefore a proton accelerator using spallation reaction with a target nucleus has an advantage to transmutation of TRU than an electron one. A fast reactor is expected to primarily have a merit that the reactor can be operated for a long term without employment of highly enriched plutonium fuel by using external neutron source such as the proton accelerator. Namely, the system has a desirable characteristic of being possible to self-sustained fissile plutonium. Consequently in the present report, core characteristics of the system were roughly studied by analyses using 2D-BURN code. The possibility of self-sustained fuel was investigated from the burnup and neutronic calculation in a cylindrical core with 300w/cc of power density without considering a target material region for the accelerator. For a reference core of which the height and the radius are both 100 cm, there is a fair prospect that a long term reactor operation is possible with subsequent refueling of natural uranium, if the medium enriched (around 10wt%) uranium or plutonium fuels are fully loaded in the initial core. More precise analyses will be planed in a later fiscal year. (author)

  16. Intelligent information database of the thermal-hydraulic characteristics for a future marine water reactor

    International Nuclear Information System (INIS)

    Inasaka, Fujio; Nariai, Hideki

    2000-01-01

    At the Ship Research Institute, a series of the experimental studies on the thermal-hydraulic characteristics of an integrated type marine water reactor has been conducted. This current study aims at developing an intelligent information database program with the thermal-hydraulic characteristics of a future marine water reactor on the basis of the valuably experimental knowledge, which was obtained from the above-mentioned studies. In this paper, the experimental knowledge with the flow boiling of a once-through steam generator and the natural circulation of primary water under a ship rolling motion was converted into an intelligent information database program. The program was created as a Windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis and determination of stability for any helical-coil type once-through steam generator design, (2) reference and graphic display of the experimental data, (3) reference of the information such as analysis method and experimental apparatus. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized reactor with helical-coil type steam generator. (author)

  17. Engineering and physics of high-power-density, compact, reversed-field-pinch fusion reactors

    International Nuclear Information System (INIS)

    Najmabadi, F.; Conn, R.W.; Krakowski, R.A.; Schultz, K.R.; Steiner, D.

    1989-01-01

    The technical feasibility and key developmental issues of compact, high-power-density Reversed-Field-Pinch (RFP) reactors are the primary results of the TITAN RFP reactor study. Two design approaches emerged, TITAN-I and TITAN-II, both of which are steady-state, DT-burning, circa 1000 MWe power reactors. The TITAN designs are physically compact and have a high neutron wall loading of 18 MW m 2 . Detailed analyses indicate that: a) each design is technically feasible; b) attractive features of compact RFP reactors can be realized without sacrificing the safety and environmental potential of fusion; and c) major features of this particular embodiment of the RFP reactor are retained in a design window of neutron wall loading ranging from 10 to 20 MW/m 2 . A major product of the TITAN study is the identification and quantification of major engineering and physics requirements for this class of RFP reactors. These findings are the focus of this paper. (author). 26 refs.; 4 figs.; 1 tab

  18. Reactor physics experiments related to transmutation in the KUCA

    Energy Technology Data Exchange (ETDEWEB)

    Shiroya, Seiji [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.

    1997-11-01

    At the Kyoto University Critical Assembly (KUCA), {sup 237}Np/{sup 235}U fission rate ratios are being measured using the back-to-back type double fission chamber to examine the nuclear data and the computational method for the transmutation of minor actinides (MA) in light water reactors (LWRs). The neutron spectra of cores are systematically being varied by changing the moderator-to-fuel volume ratio (V{sub m}/V{sub f}). The measured data are being compared with the calculated results by SRAC with three different nuclear data files. It has been indicated that the calculated results with JENDL-3.2 agreed better with the measured ones than those with JENDL-3.1 and ENDF/B-VI, although the calculated results underestimated the measured ones by around 10%. (author)

  19. 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

  20. Physics analysis of the Apollo D-3He tokamak reactor

    International Nuclear Information System (INIS)

    Santarius, J.F.; Emmert, G.A.

    1990-01-01

    Recent developments in the analysis and conceptual design of Apollo, a D- 3 He Tokamak Reactor are presented. Encouraging experimental results on TEXT motivated a key change in the Apollo concept utilization of an ergodic magnetic limiter for impurity control instead of a divertor. Parameters for the updated Apollo design and an analysis of the ergoidc magnetic limiter are given. The Apollo reference case uses direct conversion of synchrotron radiation to electricity by rectifying antennas (rectennas) for its power conversion system. Previous analyses of this concept are expanded, including further details of the rectennas and of the loss of synchrotron power to the waveguides and walls. Although Apollo will burn D- 3 He fuel, a significant amount of unburned tritium will be generated by D4D reactions. The possibility of operating a short, dedicated, T+ 3 He burn phase to eliminate this tritium will be examined