WorldWideScience

Sample records for test reactor programs

  1. The advanced test reactor strategic evaluation program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1989-01-01

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed

  2. EBR-2 [Experimental Breeder Reactor-2] test programs

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  3. EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

    International Nuclear Information System (INIS)

    Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

    1990-01-01

    The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

  4. Test program for NIS calibration to reactor thermal output in HTTR

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Shinozaki, Masayuki; Tachibana, Yukio; Kunitomi, Kazuhiko

    2000-03-01

    Rise-to-power test program for reactor thermal output measurement has been established to calibrate a neutron instrumentation system taking account of the characteristics of the High Temperature Engineering Test Reactor (HTTR). An error of reactor thermal output measurement was evaluated taking account of a configuration of instrumentation system. And the expected dispersion of measurement in the full power operation was evaluated from non-nuclear heat-up of primary coolant up to 213degC. From the evaluation, it was found that an error of reactor thermal output measurement would be less than ±2.0% at the rated power. This report presents the detailed program of rise-to-power test for reactor thermal output measurement and discusses its measurement error. (author)

  5. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

    1993-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

  6. Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.

    2008-01-01

    Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

  7. Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

    International Nuclear Information System (INIS)

    Biemiller, E.C.; Fyfitch, Stephen; Campbell, C.A.

    1994-01-01

    The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 982 o C (1800 o F)) and to simulate Regulatory Guide 1.99 database materials (austenitized at 871 o C (1600 o F)). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (University of Michigan Test Reactor) which had never been used before for this type of irradiation program. Materials taken from plate surface locations (versus 1/4 T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, are maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (260 o C and 288 o C) to determine the effect of irradiation temperature on embrittlement. (Author)

  8. A quality assurance program for nuclear power reactor materials tests at the Ford nuclear reactor

    International Nuclear Information System (INIS)

    Burn, R.R.

    1989-01-01

    The University of Michigan Nuclear Reactor Laboratory Quality Assurance Program has been established to assure that materials testing services provided to electric utilities produce accurate results in accordance with industry standards, sound engineering practice, and customer requirements. The program was prepared to comply with applicable requirements of 10CFR50, Appendix B, of the Code of Federal Regulations and a standard of the American National Standards Institute (ANSI), N45.2. The paper discusses the quality assurance program applicability, organization, qualification and training of personnel, material identification and control, examination and testing, measuring and test equipment, nonconforming test equipment, records, audits, and distribution

  9. The RERTR [Reduced Enrichment Research and Test Reactor] program:

    International Nuclear Information System (INIS)

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) program is described. After a brief summary of the results which the RERTR program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results and new developments which ocurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40 % average burnup. Good progress was made in the area of LEU usage for the production of fission 99 Mo, and in the coordination of safety evaluations related to LEU conversions of U.S. university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U 3 Si-Al with 19.75 % enrichment and U 3 Si 2 -Al with 45 % enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR program. (Author)

  10. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-15

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

  11. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  12. The Advanced Test Reactor Strategic Evaluation Program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1990-01-01

    A systematic evaluation of safety, environmental, and operational issues has been initiated at the Advanced Test Reactor (ATR). This program, the Strategic Evaluation Program (STEP), provides an integrated review of safety and operational issues against the standards applied to licensed commercial facilities. In the review of safety issues, 18 deviations were identified which required prompt attention. Resolution of these items has been accelerated in the program. An integrated living schedule is being developed to address the remaining findings. A risk evaluation is being performed on the proposed corrective actions and these actions will then be formally ranked in order of priority based on considerations of safety and operational significance. Once the final ranking is completed, an integrated schedule will be developed, which will include considerations of availability of funding and operating schedule. 3 refs., 2 figs

  13. NRC review of passive reactor design certification testing programs: Overview, progress, and regulatory perspective

    Energy Technology Data Exchange (ETDEWEB)

    Levin, A.E.

    1995-09-01

    New reactor designs, employing passive safety systems, are currently under development by reactor vendors for certification under the U.S. Nuclear Regulatory Commission`s (NRC`s) design certification rule. The vendors have established testing programs to support the certification of the passive designs, to meet regulatory requirements for demonstration of passive safety system performance. The NRC has, therefore, developed a process for the review of the vendors` testing programs and for incorporation of the results of those reviews into the safety evaluations for the passive plants. This paper discusses progress in the test program reviews, and also addresses unique regulatory aspects of those reviews.

  14. Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.

    1987-01-01

    The Argentine Atomic Energy Commission started in 1978 the Reduced Enrichment Research and Test Reactors in the field of reactor engineering; engineering, development and manufacturing of fuel elements and research reactors operators. This program was initiated with the conviction that it would contribute to the international efforts to reduce risks of nuclear weapons proliferation owing to an uncontrolled use of highly enriched uranium. It was intended to convert RA-3 reactor to make possible its operation with low enriched fuel (LEU), instead of high enriched fuel (HEU) and to develop manufacturing techniques for said LEU. Afterwards, this program was adapted to assist other countries in reactors conversion, development of the corresponding fuel elements and supply of fuel elements to other countries. (Author)

  15. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

  16. Review of Savannah River Site K Reactor inservice inspection and testing restart program

    International Nuclear Information System (INIS)

    Anderson, M.T.; Hartley, R.S.; Kido, C.

    1992-09-01

    Inservice inspection (ISI) and inservice testing (IST) programs are used at commercial nuclear power plants to monitor the pressure boundary integrity and operability of components in important safety-related systems. The Department of Energy (DOE) - Office of Defense Programs (DP) operates a Category A (> 20 MW thermal) production reactor at the Savannah River Site (SRS). This report represents an evaluation of the ISI and IST practices proposed for restart of SRS K Reactor as compared, where applicable, to current ISI/IST activities of commercial nuclear power facilities

  17. Programming for a nuclear reactor instrument simulator

    International Nuclear Information System (INIS)

    Cohn, C.E.

    1989-01-01

    A new computerized control system for a transient test reactor incorporates a simulator for pre-operational testing of control programs. The part of the simulator pertinent to the discussion here consists of two microprocessors. An 8086/8087 reactor simulator calculates simulated reactor power by solving the reactor kinetics equations. An 8086 instrument simulator takes the most recent power value developed by the reactor simulator and simulates the appropriate reading on each of the eleven reactor instruments. Since the system is required to run on a one millisecond cycle, careful programming was required to take care of all eleven instruments in that short time. This note describes the special programming techniques used to attain the needed performance

  18. NRC review of passive reactor design certification testing programs: Overview and regulatory perspective

    International Nuclear Information System (INIS)

    Levin, A.E.

    1993-01-01

    Reactor vendors are developing new designs for future deployment, including open-quotes passiveclose quotes light water reactors (LWRs), such as General Electric's (G.E.'s) simplified boiling water reactor (SBWR) and Westinghouse's AP600, which depend primarily on inherent processes, such as national convection and gravity feed, for safety injection and emergency core cooling. The U.S. Nuclear Regulatory Commission (NRC) has implemented a new process, certification of standardized reactor designs, for licensing these Plants. Part 52 of Title 10 of the Code of Federal Regulations (10CFR52) contains the requirements that vendors must meet for design certification. One important section, 10CFR52.47, reads open-quotes Certification of a standard design which . . . utilizes simplified, inherent, passive, or other innovative means to accomplish its safety functions will be granted only if: (1) The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof; (2) Interdependent effects among the safety features have been found acceptable by analysis, appropriate test programs, experience, or a combination thereof; and (3) Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses. . . . The vendors have initiated programs to test innovative features of their designs and to develop data bases needed to validate their analytical codes, as required by the design certification rule. Accordingly, the NRC is reviewing and evaluating the vendors programs to ensure that they address adequately key issues concerning safety system performance. This paper provides an overview of the NRC's review process and regulatory perspective

  19. The RERTR [Reduced Enrichment Research and Test Reactor] program: A progress report

    International Nuclear Information System (INIS)

    Travelli, A.

    1986-11-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1985, the activities, results, and new developments which occurred in 1986 are reviewed. The second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels, was expanded and its irradiation continued. Postirradiation examinations of several of these miniplates and of six previously irradiated U 3 Si 2 -Al full-size elements were completed with excellent results. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 is well under way and due for completion before the end of 1987. DOE removed an important barrier to conversions by announcing that the new LEU fuels will be accepted for reprocessing. New DOE prices for enrichment and reprocessing services were calculated to have minimal effect on HEU reactors, and to reduce by about 8 to 10% the total fuel cycle costs of LEU reactors. New program activities include preliminary feasibility studies of LEU use in DOE reactors, evaluation of the feasibility to use LEU targets for the production of fission-product 99 Mo, and responsibility for coordinating safety evaluations related to LEU conversions of US university reactors, as required by NRC. Achievement of the final program goals is projected for 1990. This progress could not have been achieved without close international cooperation, whose continuation and intensification are essential to the achievement of the ultimate goals of the RERTR Program

  20. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Quapp, W.J.; Watts, K.D.

    1985-01-01

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  1. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  2. The RERTR [Reduced Enrichment Research and Test Reactor] Program: Progress and plans

    International Nuclear Information System (INIS)

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission 99 Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U 3 Si-Al with 19.75% enrichment and U 3 Si 2 -Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program

  3. 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)

  4. Performance-based improvement of the leakage rate test program for the reactor containment of HTTR. Adoption of revised test programs containing 'Type A, Type B and Type C tests'

    International Nuclear Information System (INIS)

    Kondo, Masaaki; Emori, Koichi; Sekita, Kenji; Furusawa, Takayuki; Hayakawa, Masato; Kozawa, Takayuki; Aono, Tetsuya; Kuroha, Misao; Ouchi, Hiroshi; Kimishima, Satoru

    2008-10-01

    The reactor containment of HTTR is periodically tested to confirm leak-tight integrity by conducting overall integrated leakage rate tests, so-called 'Type A tests,' in accordance with a standard testing method provided in Japan Electric Association Code (JEAC) 4203. 'Type A test' is identified as a basic one for measuring whole leakage rates for reactor containments, it takes, however, much of cost and time of preparation, implementation and restoration of itself. Therefore, in order to upgrade the maintenance technology of HTTR, the containment leakage rate test program for HTTR was revised by adopting efficient and economical alternatives including Type B and Type C tests' which intend to measure leakage rates for containment penetrations and isolation valves, respectively. In JEAC4203-2004, following requirements are specified for adopting an alternative program: upward trend of the overall integrated leakage rate due to aging affection should not be recognized; performance criterion for combined leakage rate, that is a summation of local leakage rates evaluated by Type B and Type C tests and converted to whole leakage rates, should be established; the criterion of the combined leakage rate should be satisfied as well as of the overall integrated leakage rate; correlation between the overall integrated and combined leakage rates should be recognized. Considering the historical performances, policies of conforming to the forgoing requirements and of carrying out the revised test program were developed, which were accepted by the regulatory agency. This report presents an outline of the leakage rate tests for the reactor containment of HTTR, identifies practical issues of conventional Type A tests, and describes the conforming and implementing policies mentioned above. (author)

  5. The 'SURA' fast reactor program

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The Commissariat a l'Energie Atomique's SURA program on fast reactor safety consists of two specific testing programs on fastbreeder reactor safety: the Cabri and Scarabee programs. Both Cabri and Scarabee are examples of multinational research collaboration. The CEA and the Karlsruhe Nuclear Research Center are each covering half of the construction costs. Britain, the US and Japan are also due to participate in these experiments. The aim of the programs is to examine the behaviour of fuel in sodium cooled fast reactors. The Cabri program consists of setting off a reactivity accident in a power reactor core which is cooled with liquid sodium, such an accident occurring after a sharp increase in reactivity or as a result of the pump suddenly breaking down without there at the same time being any fall in the control rods. In 1967 the Commissariat a l'Energie Atomique started its Scarabee research program which is trying to analyse the sort of things that can go wrong with fuel cooling systems and what the consequences can be [fr

  6. Research program of the high temperature engineering test reactor for upgrading the HTGR technology

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

    1997-07-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

  7. Test reactor: basic to U.S. breeder reactor development

    International Nuclear Information System (INIS)

    Miller, B.J.; Harness, A.J.

    1975-01-01

    Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

  8. Summary of several hydraulic tests in support of the light water breeder reactor design (LWBR development program)

    International Nuclear Information System (INIS)

    McWilliams, K.D.; Turner, J.R.

    1979-05-01

    As part of the Light Water Breeder Reactor development program, hydraulic tests of reactor components were performed. This report presents the results of several of those tests performed for components which are somewhat unique in their application to a pressurized water reactor design. The components tested include: triplate orifices used for flow distribution purposes, multiventuri type flowmeters, tight lattice triangular pitch rod support grids, fuel rod end support plates, and the balance piston which is a major component of the movable fuel balancing system. Test results include component pressure loss coefficients, flowmeter coefficients and fuel rod region pressure drop characteristics

  9. The US Liquid Metal Reactor Development Program

    International Nuclear Information System (INIS)

    Till, C.E.; Arnold, W.H.; Griffith, J.D.

    1988-01-01

    The US Liquid Metal Reactor Development Program has been restructured to take advantage of the opportunity today to carry out R and D on truly advanced reactor technology. The program gives particular emphasis to improvements to reactor safety. The new directions are based on the technology of the Integral Fast Reactor (IFR). Much of the basis for superior safety performance using IFR technology has been experimentally verified and aggressive programs continue in EBR-II and TREAT. Progress has been made in demonstrating both the metallic fuel and the new electrochemical processes of the IFR. The FFTF facility is converting to metallic fuel; however, FFTF also maintains a considerable US program in oxide fuels. In addition, generic programs are continuing in steam generator testing, materials development, and, with international cooperation, aqueous reprocessing. Design studies are carried out in conjunction with the IFR technology development program. In summary, the US maintains an active development program in Liquid Metal Reactor technology, and new directions in reactor safety are central to the program

  10. Automated testing of reactor protection instrumentation made easy

    International Nuclear Information System (INIS)

    Iborra, A.; De Marcos, F.; Pastor, J.A.; Alvarez, B.; Jimenez, A.; Mesa, E.; Alsonso, L.; Regidor, J.J.

    1997-01-01

    Maintenance and testing of reactor protection systems is an important cause of unplanned reactor trips. Automated testing is the answer because it minimises test times and reduces human error. The GAMA I system, developed and implemented at Vandellos II in Spain, has the added advantage that it uses visual programming, which means that changing the software does not need specialist programming skills. (author)

  11. Fusion reactor design and technology program in China

    International Nuclear Information System (INIS)

    Huang, J.H.

    1994-01-01

    A fusion-fission hybrid reactor program was launched in 1987. The purpose of development of the hybrid reactor is twofold: to solve the problem of nuclear fuel supply for an expected large-scale development of fission reactor plants, and to maintain the momentum of fusion research. The program is described and the activities and progress of the program are presented. Two conceptual designs of an engineering test reactor with tokamak configuration were developed at the Southwestern Institute of Physics and the Institute of Plasma Physics. The results are a tokamak engineering test breeder (TETB) series design and a fusion-fission hybrid reactor design (SSEHR), characterized by a liquid-Li self-cooled blanket and an He-cooled solid tritium breeder blanket respectively. In parallel with the design studies, relevant technological experiments on a small or medium scale have been supported by this program. These include LHCD, ICRH and pellet injection in the area of plasma engineering; neutronics integral experiments with U, Pu, Fe and Be; various irradiation tests of austenitic and ferritic steels, magnetohydrodynamic (MHD) pressure drop experiments using a liquid metal loop; research into permeation barriers for tritium and hydrogen isotopes; solid tritium breeder tests using an in-situ loop in a fission reactor. All these experiments have proceeded successfully. The second step of this program is now starting. It seems reasonable that most of the research carried out in the first step will continue. ((orig.))

  12. Progress of the RERTR [Reduced Enrichment Research and Test Reactor] Program in 1989

    International Nuclear Information System (INIS)

    Travelli, A.

    1989-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs

  13. The Test Reactor Embrittlement Data Base (TR-EDB)

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.; Wang, J.A.

    1993-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is part of an ongoing program to collect test data from materials irradiations to aid in the research and evaluation of embrittlement prediction models that are used to assure the safety of pressure vessels in power reactors. This program is being funded by the US Nuclear Regulatory Commission (NRC) and has resulted in the publication of the Power Reactor Embrittlement Data Base (PR-EDB) whose second version is currently being released. The TR-EDB is a compatible collection of data from experiments in materials test reactors. These data contain information that is not obtainable from surveillance results, especially, about the effects of annealing after irradiation. Other information that is only available from test reactors is the influence of fluence rates and irradiation temperatures on radiation embrittlement. The first version of the TR-EDB will be released in fall of 1993 and contains published results from laboratories in many countries. Data collection will continue and further updates will be published

  14. IRSN research programs concerning reactor safety

    International Nuclear Information System (INIS)

    Bardelay, J.

    2005-01-01

    This paper is made up of 3 parts. The first part briefly presents the missions of IRSN (French research institute on nuclear safety), the second part reviews the research works currently led by IRSN in the following fields : -) the assessment of safety computer codes, -) thermohydraulics, -) reactor ageing, -) reactivity accidents, -) loss of coolant, -) reactor pool dewatering, -) core meltdown, -) vapor explosion, and -) fission product release. In the third part, IRSN is shown to give a major importance to experimental programs led on research or test reactors for collecting valid data because of the complexity of the physical processes that are involved. IRSN plans to develop a research program concerning the safety of high or very high temperature reactors. (A.C.)

  15. FFTF reactor-characterization program: gamma-ray measurements and shield characterization

    International Nuclear Information System (INIS)

    Bunch, W.L.; Moore, F.S. Jr.

    1983-02-01

    A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA)

  16. A program for dynamic noise investigations of reactor systems

    International Nuclear Information System (INIS)

    Antonov, N.A.; Yaneva, N.B.

    1980-01-01

    A stochastic process analysis in nuclear reactors is used for the state diagnosis and dynamic characteristic investigation of the reactor system. A program DENSITY adapted and tested on an IBM 360 ES type computer is developed. The program is adjusted for fast processing of long series exploiting a relatively small memory. The testing procedure is discussed and the method of the periodic sequences corresponding to characteristic reactivity perturbations of the reactor systems is considered. The program is written for calculating the auto-power spectral density and the cross-power spectral density, as well as the coherence function of stationary statistical time series using the advantages of the fast Fourier transformation. In particular, it is shown that the multi-frequency binary sequences are very useful with respect to the signal-to-noise ratio and the frequency distribution in view of the frequency reactor test

  17. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  18. Using low-enriched uranium in research reactors: The RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1994-01-01

    The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of 99 Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program

  19. Present status and prospects of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1995-01-01

    It is essentially important in Japan, which has limited amount of natural resources, to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors. Hence, efforts are to be continuously devoted to establish and upgrade High Temperature Gas-cooled Reactor (HTGR) technologies and to make much of research resources accumulated so far. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950degC at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). The construction of the HTTR started in March 1991, with first criticality in 1998 to be followed after commissioning testing. At present the HTTR reactor building and its containment vessel have been nearly completed and its main components, such as a reactor pressure vessel, an intermediate heat exchanger, hot gas pipings and core support structures, have been manufactured at their factories and delivered to the Oarai Research Establishment of the JAERI for their installation in the middle of 1994. Fuel fabrication will be started as well. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. The IAEA Coordinated Research Programme on Design and Evaluation of Heat Utilization Systems for the HTTR, such as steam reforming of methane and thermochemical water splitting for hydrogen production, was launched successfully in January 1994. Some heat utilization system is planned to be connected to the HTTR and demonstrated at the former stage of the second core. At present, steam-reforming of methane is the first candidate. The JAERI also plans to conduct material

  20. Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

    International Nuclear Information System (INIS)

    1975-06-01

    The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

  1. Sheath insulator final test report, TFE Verification Program

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-01

    The sheath insulator in a thermionic cell has two functions. First, the sheath insulator must electrically isolate the collector form the outer containment sheath tube that is in contact with the reactor liquid metal coolant. Second, The sheath insulator must provide for high uniform thermal conductance between the collector and the reactor coolant to remove away waste heat. The goals of the sheath insulator test program were to demonstrate that suitable ceramic materials and fabrication processes were available, and to validate the performance of the sheath insulator for TFE-VP requirements. This report discusses the objectives of the test program, fabrication development, ex-reactor test program, in-reactor test program, and the insulator seal specifications.

  2. Sheath insulator final test report, TFE Verification Program

    International Nuclear Information System (INIS)

    1994-07-01

    The sheath insulator in a thermionic cell has two functions. First, the sheath insulator must electrically isolate the collector form the outer containment sheath tube that is in contact with the reactor liquid metal coolant. Second, The sheath insulator must provide for high uniform thermal conductance between the collector and the reactor coolant to remove away waste heat. The goals of the sheath insulator test program were to demonstrate that suitable ceramic materials and fabrication processes were available, and to validate the performance of the sheath insulator for TFE-VP requirements. This report discusses the objectives of the test program, fabrication development, ex-reactor test program, in-reactor test program, and the insulator seal specifications

  3. The irradiation test program for transmutation in the French Phenix fast reactor

    International Nuclear Information System (INIS)

    Guidez, J.; Chaucheprat, P.; Fontaine, B.; Brunon, E.

    2004-01-01

    Put on commercial operation in July 1974, the French fast reactor Phenix reached a 100 000 hours operation time in september 2003. When the French law relative to long lived radioactive waste management was promulgated on December 1991, priority was given to Phenix to be run as a research reactor and to carry on a wide irradiation program dedicated to study transmutation of minor actinides and long-lived fission products. After a major renovation program required to extend the reactor lifetime, Phenix power buildup took place in 2003. Experimental irradiations have been loaded in the core, involving components for heterogeneous and homogeneous transmutation modes, americium targets, technetium 99 metal pins and isolated isotopes for integral cross-sections measurements. Associated post- irradiated examination programs are already underway or planned. With new experiments to be loaded in the core in 2006 the Phenix reactor remains to be a powerful tool providing an important experimental data on fast reactors and on transmutation of minor actinides and long-lived fission products, as well as it will contribute to gain further experience in the framework of the GENERATION IV International Forum. (authors)

  4. SP-100 reactor disassembly remote handling test program

    International Nuclear Information System (INIS)

    Wilson, C.E.; Potter, J.D.; Maiden, G.E.; Vader, D.P.

    1991-01-01

    This paper is presented as an overview of the remote handling equipment validation testing, which will be conducted before installation and use in the ground engineering test facility. This equipment will be used to defuel the SP-100 reactor core after removing it from the Test Assembly following nuclear testing. A series of full scale mock-up operational tests will be conducted at a Hanford Site facility to verify equipment design, operation, and capabilities

  5. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

    2011-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  6. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  7. Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

    International Nuclear Information System (INIS)

    Fehrenbach, P.J.; Wood, J.C.

    1987-06-01

    Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

  8. Present status of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1994-01-01

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950degC at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test using HTTR. (author)

  9. Present status of High-Temperature engineering Test Reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1993-01-01

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950 deg C at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test plan using HTTR. (author)

  10. The NRU blowdown test facility commissioning program

    Energy Technology Data Exchange (ETDEWEB)

    Walsworth, J A; Zanatta, R J; Yamazaki, A R; Semeniuk, D D; Wong, W; Dickson, L W; Ferris, C E; Burton, D H [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.

    1990-12-31

    A major experimental program has been established at the Chalk River Nuclear Laboratories (CRL) that will provide essential data on the thermal and mechanical behaviour of nuclear fuel under abnormal reactor operating conditions and on the transient release, transport and deposition of fission product activity from severely degraded fuel. A number of severe fuel damage (SFD) experiments will be conducted within the Blowdown Test Facility (BTF) at CRL. A series of experiments are being conducted to commission this new facility prior to the SFD program. This paper describes the features and the commissioning program for the BTF. A development and testing program is described for critical components used on the reactor test section. In-reactor commissioning with a fuel assembly simulator commenced in 1989 June and preliminary results are given. The paper also outlines plans for future all-effects, in-reactor tests of CANDU-designed fuel. (author). 11 refs., 3 tabs., 7 figs.

  11. Application of advanced irradiation analysis methods to light water reactor pressure vessel test and surveillance programs

    International Nuclear Information System (INIS)

    Odette, R.; Dudey, N.; McElroy, W.; Wullaert, R.; Fabry, A.

    1977-01-01

    Inaccurate characterization and inappropriate application of neutron irradiation exposure variables contribute a substantial amount of uncertainty to embrittlement analysis of light water reactor pressure vessels. Damage analysis involves characterization of the irradiation environment (dosimetry), correlation of test and surveillance metallurgical and dosimetry data, and projection of such data to service conditions. Errors in available test and surveillance dosimetry data are estimated to contribute a factor of approximately 2 to the data scatter. Non-physical (empirical) correlation procedures and the need to extrapolate to the vessel may add further error. Substantial reductions in these uncertainties in future programs can be obtained from a more complete application of available damage analysis tools which have been developed for the fast reactor program. An approach to reducing embrittlement analysis errors is described, and specific examples of potential applications are given. The approach is based on damage analysis techniques validated and calibrated in benchmark environments

  12. Development of an aging evaluation and life extension program for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Dwight, J.E. Jr.

    1988-01-01

    A life extension program has been developed for the US Department of Energy's Advanced Test Reactor. The program is an adaptation of life extension pilot programs at the Surry Unit 1 and Monticello generating stations and is being completed in three phases. In Phase 1, the critical plant components were identified. In Phase 2, existing lifetime analyses and support data for the critical components were reviewed. The results from the review give a preliminary indication that an overall plant lifetime in excess of forty years is feasible. In Phase 3, now in progress, detailed evaluations for component life extensions are being performed. 2 refs., 2 figs., 1 tab

  13. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Tehan, Terry

    2002-01-01

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  14. Summary of ORNL high-temperature gas-cooled reactor program

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-01-01

    Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

  15. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)

    1999-12-31

    In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

  16. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)

    1998-12-31

    In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

  17. The U.S. reduced enrichment research and test reactor (RERTR) program

    International Nuclear Information System (INIS)

    Travelli, A.

    1993-01-01

    Research and test reactors are widely deployed to study the irradiation behavior of materials of interest in nuclear engineering, to produce radioisotopes for medicine, industry, and agriculture, and as a basic research and teaching tool. In order to maximize neutron flux per unit power and/or to minimize capital costs and fuel cycle costs, most of these reactors were de- signed to utilize uranium with very high enrichment (in the 70% to 95% range). Research reactor fuels with such high uranium enrichment cause a potential risk of nuclear weapons proliferation. Over 140 research and test reactors of significant power (between 10 kW and 250 MW) are in operation with very highly enriched uranium in more than 35 countries, with total power in excess of 1,700 MW. The overall annual fuel requirement of these reactors corresponds to approximately 1,200 kg of 235 U. This highly strategic material is normally exported from the United States, converted to metal form, shipped to a fuel fabricator, and then shipped to the reactor site in finished fuel element form. At the reactor site the fuel is first stored, then irradiated, stored again, and eventually shipped back to the United States for reprocessing. The whole cycle takes approximately four years to complete, bringing the total required fuel inventory to approximately 5,000 kg of 235 U. The resulting international trade in highly-enriched uranium may involve several countries in the process of refueling a single reactor and creates a considerable concern that the highly-enriched uranium may be diverted for non-peaceful purposes while in fabrication, transport, or storage, particularly when it is in the unirradiated form. The proliferation resistance of nuclear fuels used in research and test reactors can be considerably improved by reducing their uranium enrichment to a value less than 20%, but significantly greater than natural to avoid excessive plutonium production

  18. Progress of the United States foreign research reactor spent nuclear fuel acceptance program. Reduced enrichment for research and test reactors conference 2002

    International Nuclear Information System (INIS)

    Clapper, Maureen

    2002-01-01

    Foreign Research Reactor Spent nuclear fuel Acceptance Program is actively working with research reactors to accept eligible material before the Acceptance Policy proper expires in 2006. Reactors/governments wishing to participate should contact US immediately if they have not done so already. Program operations are changing to adapt to new challenges. We continue to promote the importance of this Program to senior management in the Department of Energy

  19. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

  20. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  1. Results of the initial test program for the Sandia Pulsed Reactor III (SPR III)

    International Nuclear Information System (INIS)

    Estes, B.F.; Reuscher, J.A.

    1976-08-01

    This document presents a detailed discussion of the reactor including the mechanical and nuclear design characteristics. Also presented are the complete results of the Initial Approach to Critical and the Zero-and-Low Power testing programs. Reactivity worth measurements are given for such parameters as control element integral worth, Safety Block integral worth, and various materials (polyethylene, copper, lead, etc) as a function of position relative to the core. Subcritical reactivity measurements made during the approach to critical generally proved to be in reasonably good agreement with design values due to the good source-fuel-detector geometry possible with a reactor of this type. Subsequent dynamic measurements for reactivity worths are shown to be in good agreement with calculated results

  2. Ageing management program for reactor components in HANARO

    International Nuclear Information System (INIS)

    Cho, Yeong-Garp; Wu, Sang-Ik; Lee, Jung-Hee; Ryu, Jeong-Soo; Park, Yong-Chul; Wu, Jong-Sup; Jun, Byung Jin

    2003-01-01

    The HANARO, an open-tank-in-pool type research reactor of 30MWth power in Korea, has operated for 8 years since its initial criticality in February of 1995. The reactor power has been gradually increased to 24 MWth through the service period. Therefore the reactor age is very young from the viewpoint of the ageing effect on the reactor structure and components by neutron irradiation considering the expected reactor lifetime. But, we have a few programs to manage the ageing from the aspect of design lifetime of reactor components. This paper summarizes the overall progress and plan for the ageing management for the reactor components including lifetime extension and design improvement, remote measurements and in-service inspections. The shutoff units and control absorber units have aged more rapidly than other structures or components because the number of rod drop cycles was higher than expected at the design stage. The system commissioning tests, periodic performance tests, and weekly operation for the stable supply of medical radioisotopes overriding the normal cycle operation have contributed to the high frequency of rod drop. Therefore, we have instituted a program to extend the lifetime of the shutoff units and the control absorber units. This program includes an endurance test to verify the performance for the extended number of drops and the management of shutdown methods to minimize the drop cycles for both the shutoff units and the control absorber units. The program also includes the design improvement of the damper mechanism of the control absorber units to reduce the impact force caused by rod drop. The inner shell of the reflector vessel surrounding the core is the most critical part from the viewpoint of neutron irradiation. The periodic measurement of the dimensional change in the vertical straightness of the inner shell is considered as one of the in-service inspections. We developed a few tools and verified the performance to measure the

  3. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

  4. Improving nuclear safety at international research reactors: The Integrated Research Reactor Safety Enhancement Program (IRRSEP)

    International Nuclear Information System (INIS)

    Huizenga, David; Newton, Douglas; Connery, Joyce

    2002-01-01

    Nuclear energy continues to play a major role in the world's energy economy. Research and test reactors are an important component of a nation's nuclear power infrastructure as they provide training, experiments and operating experience vital to developing and sustaining the industry. Indeed, nations with aspirations for nuclear power development usually begin their programs with a research reactor program. Research reactors also are vital to international science and technology development. It is important to keep them safe from both accident and sabotage, not only because of our obligation to prevent human and environmental consequence but also to prevent corresponding damage to science and industry. For example, an incident at a research reactor could cause a political and public backlash that would do irreparable harm to national nuclear programs. Following the accidents at Three Mile Island and Chernobyl, considerable efforts and resources were committed to improving the safety posture of the world's nuclear power plants. Unsafe operation of research reactors will have an amplifying effect throughout a country or region's entire nuclear programs due to political, economic and nuclear infrastructure consequences. (author)

  5. The analysis of reactor vessel surveillance program data

    International Nuclear Information System (INIS)

    Norris, E.B.

    1979-01-01

    Commercial nuclear power reactor vessel surveillance programs are provided by the reactor supplier and are designed to meet the requirements of ASTM Method E 185. (3). Each surveillance capsule contains sets of Charpy V-notch (Csub(v)) specimens representing selected materials from the vessel beltline region and some reference steel, tension test specimens machined from selected beltline materials, temperature monitors, and neutron flux dosimeters. Surveillance capsules may also contain fracture mechanics specimens machined from selected vessel beltline materials. The major steps in the conduct of a surveillance program include (1) the testing of the surveillance specimens to determine the exposure conditions at the capsule location and the resulting embrittlement of the vessel steel, (2) the extrapolation of the capsule results to the pressure vessel wall, and (3) the determination of the heatup and cooldown limits for normal, upset, and test operation. This paper will present data obtained from commercial light water reactor surveillance programs to illustrate the methods of analysis currently in use at Southwest Research Institute and to demonstrate some of the limitations imposed by the data available. Details concerning the procedures for testing the surveillance capsule specimens will not be included because they are considered to be outside of the scope of this paper

  6. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

    International Nuclear Information System (INIS)

    1976-05-01

    Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T

  7. Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

    International Nuclear Information System (INIS)

    Wootan, D.W.; Omberg, R.P.

    2015-01-01

    Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

  8. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

  9. Training and research reactor facility longevity extension program

    International Nuclear Information System (INIS)

    Carriveau, G.W.

    1991-01-01

    Since 1943, over 550 training and research reactors have been in operation. According to statistics from the International Atomic Energy Agency, ∼325 training and research reactors are currently in service. This total includes a wide variety of designs covering a range of power and research capabilities located virtually around the world. A program has been established at General Atomics (GA) that is dedicated to the support of extended longevity of training and research reactor facilities. Aspects of this program include the following: (1) new instrumentation and control systems; (2) improved and upgraded nuclear monitoring and control channels; (3) facility testing, repair and upgrade services that include (a) pool or tank integrity, (b) cooling system, and (c) water purification system; (4) fuel element testing procedures and replacement; (5) control rod drive rebuilding and upgrades; (6) control and monitoring system calibration and repair service; (7) training services, including reactor operations, maintenance, instrumentation calibration, and repair; and (8) expanded or new uses such as neutron radiography and autoradiography, isotope production, nuclear medicine, activation analysis, and material properties modification

  10. Fiscal year 1998 multi-year work plan. Advanced reactors transition program

    International Nuclear Information System (INIS)

    Gantt, D.A.

    1997-01-01

    The mission of the Advanced Reactors Transition program is two-fold. First, the program is to maintain the Fast Flux Test Facility (FFTF) and the Fuels and Materials Examination Facility (FMEF) in Standby to support a possible future role in the tritium production strategy. Secondly, the program is to continue deactivation activities which do not conflict with the Standby directive. On-going deactivation activities include the processing of non-usable, irradiated, FFTF components for storage or disposal; deactivation of Nuclear Energy legacy test facilities; and deactivation of the Plutonium Recycle Test Reactor (PRTR) facility, 309 Building

  11. KNK I Test Program, Final Report Part 1

    International Nuclear Information System (INIS)

    Kathol, W.

    1976-01-01

    The compact sodium cooled nuclear reactor KNK I of the Karlsruhe Research Center reached full power for the first time in February 1974. The goal of KNK I is to collect experience for the construction and operation of larger reactors, such as SNR 300. In order to deepen these experiences, a test program was drawn up and conducted from 1973 until 1975 within the framework of R and D work on the development of fast breeder reactors. The program included individual tasks concerning reactor design, safety instrumentation, irradiation and post-examination as well as behavior of components during operation. The performance of the tests was essentially governed by the licensing procedure imposed under the atomic energy act for the construction and operation of nuclear facilities. This report is the first part of the final report of the test program

  12. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    International Nuclear Information System (INIS)

    Grover, S.B.

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  13. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  14. Run-Beyond-Cladding-Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    International Nuclear Information System (INIS)

    Batte, G.L.; Hoffman, G.L.

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab

  15. Nuclear Safety Research Reactor (NSRR) program in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, M; Hoshi, T; Ohnishi, N; Fujishiro, T; Inabe, T [Japan Atomic Energy Research Institute (Japan)

    1974-07-01

    An experimental research program, named Nuclear Safety Research Reactor (NSRR) Program, has been progressing in Japan Atomic Energy Research Institute (JAERI) using a modified TRIGA-ACPR. This paper is prepared to describe the outline of the NSRR program. The purpose of the NSRR program is to examine the behaviors of fuel rods under various accidental conditions of power reactors so as to establish realistic safety criteria and to develop analytical models for prediction of fuel failures. We expect to contribute finally to the improvement of reactor design and fuel fabrication techniques based on these experimental results. The NSRR experiments will be performed in the large central experimental tube, which is one of the most excellent features of this reactor, using specially designed capsules or loops which can accommodate up to 49 BWR type test fuels. Many types of test fuels in various conditions will be examined by the NSRR program, such as BWR, PWR and FBR type fuels from the beginning of life to the end of life with and without simulated reactor internal structures. The experiments will be continued for more than 10 years divided into three phases. The first phase of the program will be devoted to the experiments pertaining to reactivity initiated accidents (RIA). In these experiments we will make use of the excellent pulsing capability of ACPR, which is expected to generate 100 MW-sec prompt energy release with 1.3 msec of minimum reactor period by 4.7 dollar reactivity insertion and to yield more than 280 cal/g-UO{sub 2} heat deposit even in an approximately 10% enriched BWR type test fuel. (280 cal/g-UO{sub 2} is believed enough heat deposit to cause fuel failure.) In general, heat flow behaviors from fuel meat to clad and from clad to coolant are very complex phenomena, but they are the key point in analyzing transient response of fuels. In the sudden heat transient conditions brought by pulsing, however, it will be possible to examine each phenomenon

  16. Nuclear Safety Research Reactor (NSRR) program in JAERI

    International Nuclear Information System (INIS)

    Ishikawa, M.; Hoshi, T.; Ohnishi, N.; Fujishiro, T.; Inabe, T.

    1974-01-01

    An experimental research program, named Nuclear Safety Research Reactor (NSRR) Program, has been progressing in Japan Atomic Energy Research Institute (JAERI) using a modified TRIGA-ACPR. This paper is prepared to describe the outline of the NSRR program. The purpose of the NSRR program is to examine the behaviors of fuel rods under various accidental conditions of power reactors so as to establish realistic safety criteria and to develop analytical models for prediction of fuel failures. We expect to contribute finally to the improvement of reactor design and fuel fabrication techniques based on these experimental results. The NSRR experiments will be performed in the large central experimental tube, which is one of the most excellent features of this reactor, using specially designed capsules or loops which can accommodate up to 49 BWR type test fuels. Many types of test fuels in various conditions will be examined by the NSRR program, such as BWR, PWR and FBR type fuels from the beginning of life to the end of life with and without simulated reactor internal structures. The experiments will be continued for more than 10 years divided into three phases. The first phase of the program will be devoted to the experiments pertaining to reactivity initiated accidents (RIA). In these experiments we will make use of the excellent pulsing capability of ACPR, which is expected to generate 100 MW-sec prompt energy release with 1.3 msec of minimum reactor period by 4.7 dollar reactivity insertion and to yield more than 280 cal/g-UO 2 heat deposit even in an approximately 10% enriched BWR type test fuel. (280 cal/g-UO 2 is believed enough heat deposit to cause fuel failure.) In general, heat flow behaviors from fuel meat to clad and from clad to coolant are very complex phenomena, but they are the key point in analyzing transient response of fuels. In the sudden heat transient conditions brought by pulsing, however, it will be possible to examine each phenomenon separately

  17. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    Energy Technology Data Exchange (ETDEWEB)

    Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J B [eds.

    1995-07-01

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics.

  18. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    International Nuclear Information System (INIS)

    Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J.B.

    1995-01-01

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics

  19. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, October--December 1975

    Energy Technology Data Exchange (ETDEWEB)

    1976-05-01

    Light water reactor safety activities performed during October--December 1975 are reported. The blowdown heat transfer tests series of the Semiscale Mod-1 test program was completed. In the LOFT Program, preparations were made for nonnuclear testing. The Thermal Fuels Behavior Program completed a power-cooling-mismatch test and an irradiation effects test on PWR-type fuel rods. Model development and verification efforts of the Reactor Behavior Program included developing new analysis models for the RELAP4 computer code, subroutines for the FRAP-S and FRAP-T codes, and new models for predicting reactor fuel restructuring and zircaloy cladding behavior; an analysis of post-CHF fuel behavior was made using FRAP-T.

  20. US Advanced Light Water Reactor Program; overall objective

    International Nuclear Information System (INIS)

    Klug, N.

    1989-01-01

    The overall objective of the US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) program is to perform coordinated programs of the nuclear industry and DOE to insure the availability of licensed, improved, and simplified light water reactor standard plant designs that may be ordered in the 1990's to help meet the US electrical power demand. The discussion includes plans to meet program objectives and the design certification program. DOE is currently supporting the development of conceptual designs, configurations, arrangements, construction methods/plans, and proof test key design features for the General Electric ASBWR and the Westinghouse AP600. Key features of each are summarized. Principal milestones related to licensing of large standard plants, simplified mid-size plant development, and plant lifetime improvement are noted

  1. Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Server, W.L.; Nanstad, R.K.; Odette, G.R.

    2012-01-01

    The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

  2. United States Department of Energy breeder reactor staff training domestic program

    International Nuclear Information System (INIS)

    1984-01-01

    Two US DOE projects in the Pacific Northwest offer unique on-the-scene training opportunities at sodium-cooled fast-reactor plants: the Fast Flux Test Facility (FFTF) near Richland, Washington, which has operated successfully in a wide range of irradiation test programs since 1980; and the Experimental Breeder Reactor II (EBR-II) near Idaho Falls, Idaho, which has been in operation for approximately 20 years. Training programs have been especially designed to take advantage of this plant experience. Available courses are described

  3. Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

    International Nuclear Information System (INIS)

    Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

    1999-01-01

    The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

  4. Materials surveillance program for C-E NSSS reactor vessels

    International Nuclear Information System (INIS)

    Koziol, J.J.

    1977-01-01

    Irradiation surveillance programs for light water NSSS reactor vessels provide the means by which the utility can assess the extent of neutron-induced changes in the reactor vessel materials. These programs are conducted to verify, by direct measurement, the conservatism in the predicted radiation-induced changes and hence the operational parameters (i.e., heat-up, cooldown, and pressurization rates). In addition, such programs provide assurance that the scheduled adjustments in the operational parameters are made with ample margin for safe operation of the plant. During the past 3 years, several documents have been promulgated establishing the criteria for determining both the initial properties of the reactor vessel materials as well as measurement of changes in these initial properties as a result of irradiation. These documents, ASTM E-185-73, ''Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,'' and Appendix H to 10 CFR 50, ''Reactor Vessel Material Surveillance Program Requirements,'' are complementary to each other. They are the result of a change in the basic philosophy regarding the design and analysis of reactor vessels. In effect, the empirical ''transition temperature approach,'' which was used for design, was replaced by the ''analytical fracture mechanics approach.'' The implementation of this technique was described in Welding Research Council Bulletin 1975 and Appendix G to ASME Code Section III. Further definition of requirements appears in Appendix G to 10 CFR 50 published in July 1973. It is the intent of this paper to describe (1) a typical materials surveillance program for the reactor vessel of a Combustion Engineering NSSS, and (2) how the results of such programs, as well as experimental programs provide feed-back for improvement of materials to enhance their radiation resistance and thereby further improve the safety and reliability of future plants. (author)

  5. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    Tas, A.; Teunissen, G.

    1986-01-01

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

  6. Present status and future perspectives of research and test reactor in Japan

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Kaieda, Keisuke

    2000-01-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  7. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  8. Tests for validation of fast neutron reactors safety

    International Nuclear Information System (INIS)

    Nagata, T.; Yamashita, H.

    2001-01-01

    Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

  9. Advanced In-Pile Instrumentation for Materials Testing Reactors

    Science.gov (United States)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T. C.; Chase, B. M.; Davis, K. L.; Palmer, A. J.; Schley, R. S.

    2014-08-01

    The U.S. Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified, and the progress of other development efforts is summarized. As reported in this paper, INL researchers are currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating `advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers, are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors.

  10. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Lewis, R.A.

    1978-01-01

    Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235 U loading in the reduced-enrichment fuel elements be the same as the 235 U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant

  11. Proceedings of the international meeting on reduced enrichment for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1984-05-01

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of various programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. The papers presented during the Meeting were divided into 9 sessions and one round able discussion which concluded the Meeting. The Sessions were: Program, Fuel Development, Fuel Fabrication, Irradiation testing, Safety Analysis, Special Reactor Conversion, Reactor Design, Critical Experiments, and Reprocessing and Spent Fuel Storage. Thus, topics of this Meeting were of a very wide range that was expected to result in information exchange valuable for all the participants in the RERTR program.

  12. Proceedings of the international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1984-05-01

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of various programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. The papers presented during the Meeting were divided into 9 sessions and one round able discussion which concluded the Meeting. The Sessions were: Program, Fuel Development, Fuel Fabrication, Irradiation testing, Safety Analysis, Special Reactor Conversion, Reactor Design, Critical Experiments, and Reprocessing and Spent Fuel Storage. Thus, topics of this Meeting were of a very wide range that was expected to result in information exchange valuable for all the participants in the RERTR program

  13. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

    The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

  14. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

    2012-01-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  15. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  16. Design considerations of the irradiation test vehicle for the advanced test reactor

    International Nuclear Information System (INIS)

    Tsai, H.; Gomes, I.C.; Smith, D.L.

    1997-01-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements

  17. U.S. Domestic Reactor Conversion Programs

    International Nuclear Information System (INIS)

    Woolstenhulme, Eric

    2008-01-01

    The Conversion Projects Include: the revision of the facilities safety basis documents and supporting analysis, the fabrication of new LEU fuel, the change-out of the reactor core, and the removal of the used HEU fuel (by INL University Fuels Program or DOE-NE). The major entities involved are: the U.S. Nuclear Regulatory Commission, the University reactor department, the fuel and hardware fabricators, the Spent fuel receipt facilities, the Spent fuel shipping services, and the U.S. Department of Energy and their subcontractors. Three major Reactor Conversion Program milestones have been accomplished since 2006: the conversion of the TRIGA reactor at Texas A and M University Nuclear Science Center, the conversion of the University of Florida Training Reactor, and the conversion of the Purdue University Reactor. Four Reactor Conversion Program milestones yet to be accomplished in 2008 and 2009: the Washington State University Nuclear Radiation Center reactor, the Oregon State University TRIGA Reactor, the University of Wisconsin Nuclear Reactor, and the Neutron Radiography Reactor Facility. NNSA is committed to doing things cheaper, better, smarter, safer through a 'Lessons Learned' process. The conversion team assessed each major activity grouping: Project Initiation, Conversion Proposal Development, Fuel Fabrication and Hardware, Core Conversion, and Spent Nuclear Fuel Removal. Issues were identified and recommendations were given

  18. The PUMA test program and data analysis

    International Nuclear Information System (INIS)

    Han, J.T.; Morrison, D.L.

    1997-01-01

    The PUMA test program is sponsored by the U.S. Nuclear Regulatory Commission to provide data that are relevant to various Boiling Water Reactor phenomena. The author briefly describes the PUMA test program and facility, presents the objective of the program, provides data analysis for a large-break loss-of-coolant accident test, and compares the data with a RELAP5/MOD 3.1.2 calculation

  19. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  20. Upgrading program of the experimental fast reactor Joyo

    International Nuclear Information System (INIS)

    Yoshida, A.; Yogo, S.

    2001-01-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  1. In-service inspection program for the NCS-80 reactor pressure vessel

    International Nuclear Information System (INIS)

    Scharge, J.; Wehowsky, P.; Zeibig, H.

    1978-01-01

    The in-service inspection program of reactor pressure vessels is mainly based on the ultra-sonic method, visual checking of inner and outer surfaces as well as pressure and leak tests. The test procedure require a design of the pressure vessel suitable for the test methods and the possibility to remove the pressure vessel internals. For the outside inspection a gap of sufficient width is mandatory. The present status of the ultra-sonic method and of the inner and outer manipulators affords to conduct the in-service inspection program in form of automatic checkings. The in-service inspection program for NCS-80, the Nuclear Container-Ship design of 80,000 shp, is integrated in the refueling periods due to the request for a high availability of the ship and reactor plant

  2. 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

  3. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

  4. N Reactor Deactivation Program Plan

    International Nuclear Information System (INIS)

    Walsh, J.L.

    1993-12-01

    This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities · in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directive to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually

  5. Chinese nuclear heating test reactor and demonstration plant

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

    1992-01-01

    In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

  6. Advanced In-pile Instrumentation for Material and Test Reactors

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Unruh, T.C.; Chase, B.M.; Davis, K.L.; Palmer, A.J.; Schley, R.S.

    2013-06-01

    The US Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified; and the progress of other development efforts is summarized. As reported in this paper, INL staff is currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating 'advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors. (authors)

  7. Preliminary design studies on the Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Terry, W.J.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

    1992-08-01

    This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

  8. Current and prospective fuel test programmes in the MIR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

    2007-07-01

    MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

  9. Advanced CANDU reactor development: a customer-driven program

    International Nuclear Information System (INIS)

    Hopwood, J.M.

    2005-01-01

    The Advanced CANDU Reactor (ACR) product development program is well under way. The development approach for the ACR is to ensure that all activities supporting readiness for the first ACR project are carded out in parallel, as parts of an integrated whole. In this way design engineering, licensing, development and testing, supply chain planning, construct ability and module strategy, and planning for commissioning and operations, all work in synergy with one another. Careful schedule management :ensures that program focus stays on critical path priorities.'This paper provides an overview of the program, with an emphasis on integration to ensure maximum project readiness, This program management approach is important now that AECL is participating as the reactor vendor in Dominion Energy's DOE-sponsored Combined Construction/Operating License (COL) program. Dominion Energy selected the ACR-700 as their reference reactor technology for purposes of demonstrating the COL process. AECL's development of the ACR is unique in that pre-licensing activities are being carded out parallel in the USA and Canada, via independent, but well-communicated programs. In the short term, these programs are major drivers of ACR development. The ACR design approach has been to optimize to achieve major design objectives: capital cost reduction, robust design with ample margins, proveness by using evolutionary change from existing :reference plants, design for ease :of operability. The ACR development program maintains these design objectives for each of the program elements: Design: .Carefully selected design innovations based on the SEU fuel/light water coolant:/heavy water moderator approach. Emphasis on lessons-learned review from operating experience and customer feedback Licensing: .Safety case based on strengths of existing CANDU plus benefits of optimised design Development and Test: Choice of materials, conditions to enable incremental testing building on existing CANDU and LWR

  10. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1989-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

  11. NCSU reactor sharing program. Final technical report

    International Nuclear Information System (INIS)

    Perez, P.B.

    1997-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996

  12. Operating the Advanced Test Reactor in today's economic and regulatory environment

    International Nuclear Information System (INIS)

    Furstenau, R.V.; Patrick, M.E.; Mecham, D.C.

    1999-01-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory, is the US Department of Energy's largest and most versatile test reactor. Base programs at ATR are planned well into the 21st century. The ATR and support facilities along with an overview of current programs will be reviewed, but the main focus of the presentation will be on the impact that today's economic and regulatory concerns have had on the operation of this test reactor. Today's economic and regulatory concerns have demanded more work be completed at lower cost while increasing the margin of safety. By the beginning of the 1990 s, federal budgets for research generally and particularly for nuclear research had decreased dramatically. Many national needs continued to require testing in the ATR; but demanded lower cost, increased efficiency, improved performance, and an increased margin of safety. At the same time budgets were decreasing, there was an increase in regulatory compliance activity. The new standards imposed higher margins of safety. The new era of greater openness and higher safety standards complemented research demands to work safer, smarter and more efficiently. Several changes were made at the ATR to meet the demands of the sponsors and public. Such changes included some workforce reductions, securing additional program sponsors, upgrading some facilities, dismantling other facilities, and implementing new safety programs. (author)

  13. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately

  14. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  15. U.S. uranium supply to the research and test reactor community

    International Nuclear Information System (INIS)

    Parker, Elaine M.

    2002-01-01

    From the 1950s through the early 1990s, the U.S. Department of Energy (DOE) was the primary supplier of low enriched uranium (LEU) and highly enriched uranium (HEU) to research and test reactors worldwide. The formerly called Y-12 Plant in Oak Ridge, Tennessee, was put into operational stand down in 1994 due to inadequate safety documentation. This paper will discuss the re-start of the Y-12 Plant and its current capabilities. Additionally, the paper will address recent changes within the DOE, with the creation of the National Nuclear Security Administration (NNSA). It will show how the change to NNSA and an organizational re-alignment has improved efficiencies. NNSA is committed to operate its sales program so that it is complementary to, and in support of, the Reduced Enrichment for Research and Test Reactors (RERTR) and Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Return Programs. The NNSA is committed to provide an assurance of competitively-priced, high-quality uranium supply to the research and test reactor community under long-term contracts. This paper will discuss some of NNSA's recent successes in long-term contracting and meeting deliveries. (author)

  16. Use of multiple on-campus reactors in education and training programs

    International Nuclear Information System (INIS)

    Schlapper, G.A.

    1989-01-01

    In its undergraduate and graduate programs in nuclear engineering and health physics, Texas A ampersand M University utilizes two reactors for the training and education of students. The 5-W AGN-201 nuclear training reactor has been in use since the late 1950s, while the 1-MW TRIGA Nuclear Science Center Reactor (NSCR) was first utilized in late 1961. Both facilities have been upgraded since initial criticality, the AGN power level being increased from the original 200-mW limit to its 5-W current level and the NSCR undergoing conversion from a 100-kW materials test reactor fueled deign to a 1-MW TRIGA-fueled facility. The AGN reactor is operated by the Department of Nuclear Engineering of the College of Engineering and is almost solely utilized in training and education programs. The NSCR facility is administered by the Texas Engineering Experiment Station and support research efforts of faculty and students of departments within and outside the university in addition to contributing to the education and training programs of the nuclear engineering department

  17. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

  18. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo

    2013-06-01

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  19. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Calabrese, Carlos R.

    2001-01-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  20. The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

    2001-07-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

  1. Seismic behaviour of LMFBR reactor cores. The SYMPHONY program

    International Nuclear Information System (INIS)

    Broc, Daniel

    2001-01-01

    As part of a comprehensive program on the seismic behaviour of the LMFBR reactor cores, the SYMPHONY experimental program, performed at the CEA Saclay, is carried out from 1993 up to now. LMFBR reactor cores are composed of fuel assemblies and neutronic shields, immersed in sodium (the primary coolant) or water (for the experimental tests). The main objective of the seismic studies is to evaluate the assembly motions, with consequences on the reactivity and the control rod insertability, and to verify the structural integrity of the assemblies under the impact forces. The experimental program has reached its objectives. Tests have been performed in a satisfying way. Instrumentation allowed to collect displacements, accelerations, and shock forces. All the results constitute a comprehensive base of valuable and reliable data. The interpretation of the tests is based on beam models, taking into account the Fluid Structure Interaction, and the shocks between the assemblies. Theoretical results are in a quite good agreement with the experimental ones. The interpretation of the hexagonal tests in water pointed out very strong coupling between the assemblies and lead to the development of a specific Fluid Structure Interaction, taking into account not only inertial effects, but dissipative effects also. (author)

  2. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  3. The in-core experimental program at the MIT Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kohse, G.E.; Hu, L-W., E-mail: kohse@mit.edu [Massachusetts Inst. of Technology, Nuclear Reactor Lab., Cambridge, Massachusetts (United States)

    2014-07-01

    This paper describes the program of in-core experiments at the Massachusetts Institute of Technology Research Reactor (MITR), a 6 MW research reactor. The MITR has a neutron flux and spectrum similar to those in water-cooled power reactors and therefore provides a useful test environment for materials and fuels research. In-core facilities include: a water loop operating at pressurized water or boiling water reactor conditions, an inert gas irradiation facility operating at temperature up to 850 {sup o}C and special purpose facilities including fuel irradiation experiments. Recent and ongoing tests include: water loop investigations of corrosion and thermal and mechanical property evolution of SiC/SiC composites for fuel cladding, irradiation of advanced materials and in-core sensors at elevated temperatures, irradiation in molten fluoride salt at 700 {sup o}C of metal alloy, graphite and composite materials for power reactor applications and instrumented irradiations of metal-bonded hydride fuel. (author)

  4. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

  5. Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

    2002-08-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

  6. Fast-reactor-data testing of ENDF/B-V at ORNL

    International Nuclear Information System (INIS)

    Wright, R.Q.; Ford, W.E. III; Lucius, J.L.; Webster, C.C.; Marable, J.H.

    1982-01-01

    The Cross Section Evaluation Working Group (CSEWG) is coordinating a program to assess the adequacy of ENDF/B-V cross sections for both fast- and thermal-reactor design applications. A secondary goal is to evaluate cross-section processing codes, cross-section libraries, and radiation-transport codes. Fast reactor data testing (FRDT) goals are accomplished, in part, by comparison of calculated results with documented performance parameters of CSEWG fast reactor benchmarks and with results obtained by other data testers. The purpose of this paper is to describe the results of FRDT at Oak Ridge National Laboratory

  7. Trends in large-scale testing of reactor structures

    International Nuclear Information System (INIS)

    Blejwas, T.E.

    2003-01-01

    Large-scale tests of reactor structures have been conducted at Sandia National Laboratories since the late 1970s. This paper describes a number of different large-scale impact tests, pressurization tests of models of containment structures, and thermal-pressure tests of models of reactor pressure vessels. The advantages of large-scale testing are evident, but cost, in particular limits its use. As computer models have grown in size, such as number of degrees of freedom, the advent of computer graphics has made possible very realistic representation of results - results that may not accurately represent reality. A necessary condition to avoiding this pitfall is the validation of the analytical methods and underlying physical representations. Ironically, the immensely larger computer models sometimes increase the need for large-scale testing, because the modeling is applied to increasing more complex structural systems and/or more complex physical phenomena. Unfortunately, the cost of large-scale tests is a disadvantage that will likely severely limit similar testing in the future. International collaborations may provide the best mechanism for funding future programs with large-scale tests. (author)

  8. University Reactor Instrumentation Program

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1992-11-01

    Recognizing that the University Reactor Instrumentation Program was developed in response to widespread needs in the academic community for modernization and improvement of research and training reactors at institutions such as the University of Florida, the items proposed to be supported by this grant over its two year period have been selected as those most likely to reduce foreed outages, to meet regulatory concerns that had been expressed in recent years by Nuclear Regulatory Commission inspectors or to correct other facility problems and limitations. Department of Energy Grant Number DE-FG07-90ER129969 was provided to the University of Florida Training Reactor(UFTR) facility through the US Department of Energy's University Reactor Instrumentation Program. The original proposal submitted in February, 1990 requested support for UFTR facility instrumentation and equipment upgrades for seven items in the amount of $107,530 with $13,800 of this amount to be the subject of cost sharing by the University of Florida and $93,730 requested as support from the Department of Energy. A breakdown of the items requested and total cost for the proposed UFTR facility instrumentation and equipment improvements is presented

  9. Testing of the multi-application small light water reactor (MASLWR) passive safety systems

    International Nuclear Information System (INIS)

    Reyes, Jose N.; Groome, John; Woods, Brian G.; Young, Eric; Abel, Kent; Yao, You; Yoo, Yeon Jong

    2007-01-01

    Experimental thermal hydraulic research has been conducted at Oregon State University for the purpose of assessing the performance of a new reactor design concept, the multi-application small light water reactor (MASLWR). The MASLWR is a pressurized light water reactor design with a net output of 35 MWe that uses natural circulation in both normal and transient operation. Due to its small size, portability and modularity, the MASLWR design is well suited to help fill the potential need for grid appropriate reactor designs for smaller electricity grids as may be found in developing or remote regions. The purpose of the OSU MASLWR test facility is to assess the operation of the MASLWR under normal full operating pressure and full temperature conditions and to assess the passive safety systems under transient conditions. The data generated by the testing program will be used to assess computer code calculations and to provide a better understanding of the thermal-hydraulic phenomena in the design of the MASLWR NSSS. During this testing program, four tests were conducted at the OSU MASLWR test facility. These tests included one design basis accident and one beyond design basis accident. During the performance of these tests, plant operations to include start up, normal operation and shut down evolutions were demonstrated successfully

  10. Irradiation facilitates at the advanced test reactor

    International Nuclear Information System (INIS)

    Grover, Blaine S.

    2006-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

  11. Off reactor testings. Technological engineering applicative research

    International Nuclear Information System (INIS)

    Doca, Cezar

    2001-01-01

    By the end of year 2000 over 400 nuclear electro-power units were operating world wide, summing up a 350,000 MW total capacity, with a total production of 2,300 TWh, representing 16% of the world's electricity production. Other 36 units, totalizing 28,000 MW, were in construction, while a manifest orientation towards nuclear power development was observed in principal Asian countries like China, India, Japan and Korea. In the same world's trend one find also Romania, the Cernavoda NPP Unit 1 generating electrical energy into the national system beginning with 2 December 1996. Recently, the commercial contract was completed for finishing the Cernavoda NPP Unit 2 and launching it into operation by the end of year 2004. An important role in developing the activity of research and technological engineering, as technical support for manufacturing the CANDU type nuclear fuel and supplying with equipment the Cernavoda units, was played by the Division 7 TAR of the INR Pitesti. Qualification testings were conducted for: - off-reactor CANDU type nuclear fuel; - FARE tools, pressure regulators, explosion proof panels; channel shutting, as well as functional testing for spare pushing facility as a first step in the frame of the qualification tests for the charging/discharging machine (MID) 4 and 5 endings. Testing facilities are described, as well as high pressure hot/cool loops, measuring chains, all of them fulfilling the requirements of quality assurance. The nuclear fuel off-reactor tests were carried out to determine: strength; endurance; impact, pressure fall and wear resistance. For Cernavoda NPP equipment testings were carried out for: the explosion proof panels, pressure regulators, behaviour to vibration and wear of the steam generation tubings, effects of vibration upon different electronic component, channel shutting (for Cernavoda Unit 2), MID operating at 300 and 500 cycles. A number of R and D programs were conducted in the frame of division 7 TAR of INR

  12. Correlations between power and test reactor data bases

    International Nuclear Information System (INIS)

    Guthrie, G.L.; Simonen, E.P.

    1989-02-01

    Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

  13. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1980-01-01

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  14. Study on the seismic verification test program on the experimental multi-purpose high-temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Taketani, K.; Aochi, T.; Yasuno, T.; Ikushima, T.; Shiraki, K.; Honma, T.; Kawamura, N.

    1978-01-01

    The paper describes a program of experimental research necessary for qualitative and quantitative determination of vibration characteristics and aseismic safety on structure of reactor core in the multipurpose high temperature gas-cooled experimental reactor (VHTR Experimental Reactor) by the Japan Atomic Energy Research Institute

  15. The Tokamak Fusion Test Reactor D-T modifications and operations

    International Nuclear Information System (INIS)

    1992-01-01

    This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

  16. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  17. Nuclear Power Reactor simulator - based training program

    International Nuclear Information System (INIS)

    Abdelwahab, S.A.S.

    2009-01-01

    nuclear power stations will continue playing a major role as an energy source for electric generation and heat production in the world. in this paper, a nuclear power reactor simulator- based training program will be presented . this program is designed to aid in training of the reactor operators about the principles of operation of the plant. also it could help the researchers and the designers to analyze and to estimate the performance of the nuclear reactors and facilitate further studies for selection of the proper controller and its optimization process as it is difficult and time consuming to do all experiments in the real nuclear environment.this program is written in MATLAB code as MATLAB software provides sophisticated tools comparable to those in other software such as visual basic for the creation of graphical user interface (GUI). moreover MATLAB is available for all major operating systems. the used SIMULINK reactor model for the nuclear reactor can be used to model different types by adopting appropriate parameters. the model of each component of the reactor is based on physical laws rather than the use of look up tables or curve fitting.this simulation based training program will improve acquisition and retention knowledge also trainee will learn faster and will have better attitude

  18. The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Vesterlund, G.; Vaernild, O.

    1980-01-01

    The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

  19. UCLA program in reactor studies: The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on ''modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D- 3 He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs

  20. The United States fluoride-salt-cooled high-temperature reactor program

    International Nuclear Information System (INIS)

    Holcomb, David E.

    2013-01-01

    ) is leading an integrated university research project team consisting of the University of California at Berkeley (UCB) and the University of Wisconsin (UW). The university program includes developing a test reactor conceptual design, hydraulic and thermal performance modeling using surrogate materials, and material performance testing. Ohio State University, Georgia Institute of Technology, and Johns Hopkins University also have FHR supportive technology development programs under way. (author)

  1. 'Experience with decommissioning of research and test reactors at Argonne National Laboratory'

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Yule, T.J.; Fellhauer, C.R.; Boing, L.E.

    2002-01-01

    A large number of research reactors around the world have reached the end of their useful operational life. Many of these are kept in a controlled storage mode awaiting decontamination and decommissioning (D and D). At Argonne National Laboratory located near Chicago in the United States of America, significant experience has been gained in the D and D of research and test reactors. These experiences span the entire range of activities in D and D - from planning and characterization of the facilities to the eventual disposition of all waste. A multifaceted D nd D program has been in progress at the Argonne National Laboratory - East site for nearly a decade. The program consists of three elements: - D and D of nuclear facilities on the site that have reached the end of their useful life; - Development and demonstrations of technologies that help in safe and cost effective D and D; - Presentation of training courses in D and D practices. Nuclear reactor facilities have been constructed and operated at the ANL-E site since the earliest days of nuclear power. As a result, a number of these early reactors reached end-of-life long before reactors on other sites and were ready for D and D earlier. They presented an excellent set of test beds on which D and D practices and technologies could be demonstrated in environments that were similar to commercial reactors, but considerably less hazardous. As shown, four reactor facilities, plutonium contaminated glove boxes and hot cells, a cyclotron facility and assorted other nuclear related facilities have been decommissioned in this program. The overall cost of the program has been modest relative to the cost of comparable projects undertaken both in the U.S. and abroad. The safety record throughout the program was excellent. Complementing the actual operations, a set of D and D technologies are being developed. These include robotic methods of tool handling and operation, chemical and laser decontamination techniques, sensors

  2. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

  3. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    International Nuclear Information System (INIS)

    Stephens, S.V.

    1976-09-01

    The objective of the document is to present in one volume an overview of the Argonne National Laboratory test facilities involved in the conduct of the national LMFBR research and development program. Existing facilities and those under construction or authorized as of September 1976 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. The volume is divided into two sections: Argonne-East and Argonne-West. Introductory material for each section includes site and facility maps. The profiles are arranged alphabetically by title according to their respective locations at Argonne-East or Argonne-West. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc., involved in the LMFBR program is appended

  4. Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

    Science.gov (United States)

    Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

    2017-12-01

    A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

  5. Containment Leakage Rate Testing Program in NPP Krsko

    International Nuclear Information System (INIS)

    Dudas, M.; Heruc, Z.

    2002-01-01

    NPP Krsko adopted new regulations for testing of the reactor building containment as stipulated by 10CFR50 (Code of Federal Regulation) Appendix J, Option B instead of the previous requirement 10CFR50 Appendix J now renamed to 10CFR50 Appendix J, Option A. In the USA a thorough analyses of nuclear power plants reactor building containment testing was conducted. As part of these analyses the test results obtained from testing of various reactor-building containments in the last ten years were reviewed. It was concluded that it would be meaningful to, based on test results historical data, reconsider possibility of redefining testing intervals. The official proposal of such approach was reviewed and approved by the NRC and published in September of 1995 in the FR Vol.60 No.186. Based on directions from 10CFR50 Appendix J, Option B, the new criteria for definition of test intervals were created. Criteria were based upon past performance during testing (Performance-based Requirements) and safety impact. At NPP Krsko, the analyses of the Reactor Building Containment. Integrity Test results was performed . This included test results of the Containment Integrated Leak Rate Testing (CILRT or Type A tests), Containment Isolation Valves Local Leak Rater Tests (Type C tests) and Mechanical and Electrical Penetrations Local Leak Rate tests (Type B tests). In accordance with instructions from NEI 94-01 and based on analyses of test results, NPP Krsko created Containment Leakage Rate Testing Program with the purpose to establish the performance-based definition of test intervals, inspection scope, trending and reporting. Equally, the program gives instructions how to evaluate test results and how to deal with the containment penetration or isolation valve repair contingency. All changes caused with transition from Option A to Option B are marginal to public safety. (author)

  6. Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

    International Nuclear Information System (INIS)

    Cowell, B.S.

    1997-06-01

    This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy

  7. LOCA simulation in the NRU reactor: materials test-1

    International Nuclear Information System (INIS)

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607 0 F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions

  8. The United States Advanced Reactor Technologies Research and Development Program

    International Nuclear Information System (INIS)

    O’Connor, Thomas J.

    2014-01-01

    The following aspects are addressed: • Nuclear energy mission; • Reactor research development and deployment (RD&D) programs: - Light Water Reactor Sustainability Program; - Small Modular Reactor Licensing Technical Support; - Advanced Reactor Technologies (ART)

  9. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    International Nuclear Information System (INIS)

    Loika, E.F.

    1994-01-01

    This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

  10. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  11. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  12. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  13. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  14. Summary remarks and recommended reactions for an international data file for dosimetry applications for LWR, FBR, and MFR reactor research, development and testing programs

    International Nuclear Information System (INIS)

    McElroy, W.N.; Lippincott, E.P.; Grundl, J.A.; Fabry, A.; Dierckx, R.; Farinelli, U.

    1979-01-01

    The need for the use of an internationally accepted data file for dosimetry applications for light water reactor (LWR), fast breeder reactor (FBR), and magnetic fusion reactor (MFR) research, development, and testing programs continues to exist for the Nuclear Industry. The work of this IAEA meeting, therefore, will be another important step in achieving consensus agreement on an internationally recommended file and its purpose, content, structure, selected reactions, and associated uncertainy files. Summary remarks and a listing of recommended reactions for consideration in the formulation of an ''International Data File for Dosimetry Applications'' are presented in subsequent sections of this report

  15. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    International Nuclear Information System (INIS)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes

  16. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

  17. Review of inservice inspection and nondestructive examination practices at DOE Category A test and research reactors

    International Nuclear Information System (INIS)

    Anderson, M.T.; Aldrich, D.A.

    1990-09-01

    In-service inspection (ISI) programs are used at commercial nuclear power plants for monitoring the pressure boundary integrity of various systems and components to ensure their continued safe operation. The Department of Energy (DOE) operates several test and research reactors. This report represents an evaluation of the ISI and nondestructive examination (NDE) practices at five DOE Category A (> 20 MW thermal) reactors as compared, where applicable, to the current ISI activities of commercial nuclear power facilities. The purpose of an inservice inspection (ISI) program is to establish regular surveillance of safety-related components to ensure their safe and reliable operation. The integrity of materials comprising these components is generally monitored by means of periodic nondestructive examinations (NDE), which, if appropriately performed, provide methods for identifying degradation that could render components unable to perform their intended safety functions. The reactors evaluated during this review were the Experimental Breeder Reactor 2 and the Fast Flux Test Facility (liquid-metal cooled plants), the Advanced Test Reactor and the High Flux Isotopes Reactor (light-water cooled reactors), and the High Flux Beam Reactor (a heavy-water cooled facility). Although these facilities are extremely diverse in design and operation, they all have less stored energy, smaller inventories of radionuclides, and generally, more remote locations than commercial reactors. However, all DOE test and research facilities contain components similar to those of commercial reactors for which continued integrity is important to maintain plant safety. 10 refs., 6 tabs

  18. Development of inspection and maintenance program for reactor and reactivity control units in HANARO

    International Nuclear Information System (INIS)

    Cho, Yeong-Garp

    1998-01-01

    This paper summarizes the overall program for inspection and maintenance of reactor structure and Reactivity Control Units (RCU) of HANARO during lifetime. The long-term plan for in-service inspection is introduced in the viewpoint of the structural integrity of reactor and RCU, and the operability of RCU mechanism. This program includes the list of components to be inspected, the schedule of inspection and maintenance, and the development of special tools and test rig that are required for the remote inspection and maintenance of reactor and RCU components. Preliminary results of the evaluation on the lifetime of RCU components are summarized based on the operation history since the installation of reactor. A test rig will be designed and constructed for the purposes of verifying the prolonged lifetime of RCU components being used, the performance of special tools, and the rehearsal of maintenance work as well. (author)

  19. PBF/LOFT Lead Rod Test Program experiment operating specification

    International Nuclear Information System (INIS)

    Varacalle, D.J. Jr.

    1978-11-01

    The PBF/LOFT Lead Rod (LLR) Test Program is being conducted to provide experimental information on the behavior of nuclear fuel under normal and accident conditions in the Power Burst Facility at the Idaho National Engineering Laboratory. Understanding the behavior of light-water reactors (LWR) under loss-of-coolant conditions is a major objective of the NRC Reactor Safety Research Program. The Loss of Fluid Test (LOFT) facility is the major testing facility to evaluate the systems response of an LWR over a wide range of Loss of Coolant Experment (LOCE) conditions. As such, the LOFT core is intended to be used for sequential LOCE tests provided no significant fuel rod failures occur. The PFB/LLR tests are designed to simulate the test conditions for the LOFT Power Ascension Tests L2-2 through L2-5. The test program has been designed to provide a parametric evaluation of the LOFT fuel over a wide range of power. Thus, a relatively accurate assessment of the state of the LOFT core after the completion of each subtest and the anticipated effect of the next test can be obtained by utilizing a combination of LLR test data and analytical predictions. Specifications for the test program are presented

  20. Program summary for the Civilian Reactor Development Program

    International Nuclear Information System (INIS)

    1982-07-01

    This Civilian Reactor Development Program document has the prime purpose of summarizing the technical programs supported by the FY 1983 budget request. This section provides a statement of the overall program objectives and a general program overview. Section II presents the technical programs in a format intended to show logical technical interrelationships, and does not necessarily follow the structure of the formal budget presentation. Section III presents the technical organization and management structure of the program

  1. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

  2. Scheduling and recording of reactor maintenance and testing by computer

    International Nuclear Information System (INIS)

    Gray, P.L.

    1975-01-01

    The use of a computer program, Maintenance Information and Control (MIAC), at the Savannah River Laboratory (SRL) assists a small operating staff in maintaining three research reactors and a subcritical facility. The program schedules and defines preventive maintenance, schedules required periodic tests, logs repair and cost information, specifies custodial and service responsibilities, and provides equipment maintenance history, all with a minimum of record-keeping

  3. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  4. Reactor use in nuclear engineering programs

    International Nuclear Information System (INIS)

    Murray, R.L.

    1975-01-01

    Nuclear reactors for dual use in training and research were established at about 50 universities in the period since 1950, with assistance by the U. S. Atomic Energy Commission and the National Science Foundation. Most of the reactors are in active use for a variety of educational functions--laboratory teaching of undergraduates and graduate students, graduate research, orientation of visitors, and nuclear power plant reactor operator training, along with service to the technical community. As expected, the higher power reactors enjoy a larger average weekly use. Among special programs are reactor sharing and high-school teachers' workshops

  5. Program of RA reactor start-up to nominal power

    International Nuclear Information System (INIS)

    1959-01-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is described in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members

  6. Research reactors and materials testing

    International Nuclear Information System (INIS)

    Vidal, H.

    1986-01-01

    Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

  7. Developmental Light-Water Reactor Program

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-12-01

    This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

  8. Foreign research reactor uranium supply program: The Y-12 national security complex process

    International Nuclear Information System (INIS)

    Nelson, T.; Eddy, B.G.

    2010-01-01

    The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the HEU Disposition Program, the Reduced Enrichment Research and Test Reactors (RERTR) Program, and the United States FRR Spent Nuclear Fuel (SNF) Acceptance Program. The Y-12 National Nuclear Security Administration (NNSA) Y-12 Site Office maintains the prime contracts with foreign governments for the supply of Low-Enriched Uranium (LEU) for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. This program supports the important U.S. government and nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to LEU fuel under the guidance of the NNSA RERTR Program. In conjunction with the FRR SNF Acceptance Program which supports the global nonproliferation efforts to disposition U.S.-origin HEU, the Y-12 FRR Uranium Supply Program can provide the LEU for the replacement fuel fabrication. In addition to feedstock for fuel fabrication, Y-12 supplies LEU for target fabrication for medical isotope production. The Y-12 process uses supply forecasting tools, production improvements and efficient delivery preparations to successfully support the global research reactor community

  9. The SPHINX reactor for engineering tests

    International Nuclear Information System (INIS)

    Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

    1990-01-01

    A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

  10. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, April--June 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-08-01

    Results from the previously conducted Semiscale Mod-1 ECC injection test series were analyzed. Testing in the LOFT counterpart test series was essentially completed, and the steam generator tube rupture test series was begun. Two tests in the alternate ECC injection test series were conducted which included injection of emergency core coolant into the upper plenum through use of the low pressure injection system. The Loss-of-Fluid Test Program successfully completed nonnuclear Loss-of-Coolant Experiment L1-4. A nuclear test, GC 2-3, in the Power Burst Facility Reactor was performed to evaluate the power oscillation method of determining gap conductance and to determine the effects of initial gap size, fill gas composition, and fuel density on the thermal performance of a light water reactor fuel rod. Additional test results were obtained relative to the behavior of irradiated fuel rods during a fast power increase and during a high power film boiling transient. Fuel model development and verification activities continued for the steady state and transient Fuel Rod Analysis Program, FRAP-S and FRAP-T. A computer code known as RELAP4/MOD7 is being developed to provide best-estimate modeling for reflood during a postulated loss-of-coolant accident (LOCA). A prediction of the fourth test in the boiling water reactor (BWR) Blowdown/Emergency Core Cooling Program was completed and an uncertainty analysis was completed of experimental steady state stable film boiling data for water flowing vertically upward in round tubes. A new multinational cooperative program to study the behavior of entrained liquid in the upper plenum and cross flow in the core during the reflood phase of a pressurized water reactor LOCA was defined.

  11. Development of Research Reactor Simulator and Its Application to Dynamic Test-bed

    International Nuclear Information System (INIS)

    Kwon, Kee Choon; Park, Jae Chang; Lee, Seung Wook; Bang, Dane; Bae, Sung Won

    2014-01-01

    We developed HANARO and the Jordan Research and Training Reactor (JRTR) real-time simulator for operating staff training. The main purpose of this simulator is operator training, but we modified this simulator as a dynamic test-bed to test the reactor regulating system in HANARO or JRTR before installation. The simulator configuration is divided into hardware and software. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The simulator software is divided into three major parts: a mathematical modeling module, which executes the plant dynamic modeling program in real-time, an instructor station module that manages user instructions, and a human machine interface (HMI) module. The developed research reactors are installed in the Korea Atomic Energy Research Institute nuclear training center for reactor operator training. To use the simulator as a dynamic test-bed, the reactor regulating system modeling software of the simulator was replaced by a hardware controller and the simulator and target controller were interfaced with a hard-wired and network-based interface

  12. Development of Research Reactor Simulator and Its Application to Dynamic Test-bed

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee Choon; Park, Jae Chang; Lee, Seung Wook; Bang, Dane; Bae, Sung Won [KAERI, Daejeon (Korea, Republic of)

    2014-08-15

    We developed HANARO and the Jordan Research and Training Reactor (JRTR) real-time simulator for operating staff training. The main purpose of this simulator is operator training, but we modified this simulator as a dynamic test-bed to test the reactor regulating system in HANARO or JRTR before installation. The simulator configuration is divided into hardware and software. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The simulator software is divided into three major parts: a mathematical modeling module, which executes the plant dynamic modeling program in real-time, an instructor station module that manages user instructions, and a human machine interface (HMI) module. The developed research reactors are installed in the Korea Atomic Energy Research Institute nuclear training center for reactor operator training. To use the simulator as a dynamic test-bed, the reactor regulating system modeling software of the simulator was replaced by a hardware controller and the simulator and target controller were interfaced with a hard-wired and network-based interface.

  13. New Production Reactors Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    1990-12-01

    Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs.

  14. New Production Reactors Program Plan

    International Nuclear Information System (INIS)

    1990-12-01

    Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. The development of fast simulation program for marine reactor parameters

    International Nuclear Information System (INIS)

    Chen Zhiyun; Hao Jianli; Chen Wenzhen

    2012-01-01

    Highlights: ► The simplified physical and mathematical models are proposed for a marine reactor system. ► A program is developed with Simulink module and Matlab file. ► The program developed has the merit of easy input preparation, output processing and fast running. ► The program can be used for the fast simulation of marine reactor parameters on the operating field. - Abstract: The fast simulation program for marine reactor parameters is developed based on the Simulink simulating software according to the characteristics of marine reactor with requirement of maneuverability and acute and fast response. The simplified core physical and thermal model, pressurizer model, steam generator model, control rod model, reactivity model and the corresponding Simulink modules are established. The whole program is developed by coupling all the Simulink modules. Two typical transient processes of marine reactor with fast load increase at low power level and load rejection at high power level are adopted to verify the program. The results are compared with those of Relap5/Mod3.2 with good consistency, and the program runs very fast. It is shown that the program is correct and suitable for the fast and accurate simulation of marine reactor parameters on the operating field, which is significant to the marine reactor safe operation.

  17. Test reactor risk assessment methodology

    International Nuclear Information System (INIS)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

  18. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1980-September 30, 1980

    International Nuclear Information System (INIS)

    1980-01-01

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, 950 and 1050 0 C. Initiation of controlled purity helium creep-rupture testing in the intensive screening test program is discussed. In addition, the results of 1000-hour exposures at 750 and 850 0 C on several experimental alloys are discussed

  19. Materials Inventory Database for the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  20. Safety analysis calculations for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

    1983-08-01

    The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

  1. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Benson, J.B.; Foster, J.A.; Marshall, F.M.; Meyer, M.K.; Thelen, M.C.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  2. Test reactors in the world

    International Nuclear Information System (INIS)

    Corella, M.R.; Gomez Alonso, M.

    1983-01-01

    INFCE work on research reactor core conversion from HEU to LEU, attracted a raising interest on this type of nuclear reactors. In this context, the present work shows a compilation of worldwide research and test nuclear reactors, now in operation, under construction, or planned, as well as decommissioned reactors (tables A to F). Brief descriptions of these reactors are included in tables G to L. In table M a summary view of reactors with power level between 10 and 30 MWt is shown. Attention is focused on that power range, as it has been considered in very preliminar studies for a new research reactor. Almost all data have been obtained from current available bibliography. (author)

  3. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

    1992-05-01

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  4. SP-100 Program: space reactor system and subsystem investigations

    International Nuclear Information System (INIS)

    Harty, R.B.

    1983-01-01

    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs

  5. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  6. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm 3 was by then in routine use, illustrated how far work has progressed

  7. The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

    International Nuclear Information System (INIS)

    Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

    1992-07-01

    This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

  8. Fusion reactor materials program plan. Section 2. Damage analysis and fundamental studies

    International Nuclear Information System (INIS)

    1978-07-01

    The scope of this program includes: (1) Development of procedures for characterizing neutron environments of test facilities and fusion reactors, (2) Theoretical and experimental investigations of the influence of irradiation environment on damage production, damage microstructure evolution, and mechanical and physical property changes, (3) Identification and, where appropriate, development of essential nuclear and materials data, and (4) Development of a methodology, based on damage mechanisms, for correlating the mechanical behavior of materials exposed to diverse test environments and projecting this behavior to magnetic fusion reactor (MFR) environments. Some major problem areas are addressed

  9. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1992-01-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  10. A review of fast reactor program in Japan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    In accordance with the Long-term Program for Development and Utilization of Nuclear Energy defined by the Japan Atomic Energy Commission (JAEC), Power Reactor and Nuclear Fuel Development Corporation (PNC) is playing the key role in the development of a plutonium utilization system by fast breeder reactor (FBR), which is superior to the uranium utilization system by light water reactor, aiming to achieve future stable long-term energy supply and energy security of Japan. The experimental reactor Joyo, located in the O-arai Engineering Center (OEC) of PNC, has provided abundant experimental data and excellent operational records attaining 43,500 hours operation in total by the end of 1991, since its first criticality in 1977. On the prototype reactor Monju, 97.6% of construction works has already been completed and the function tests are in progress aiming at the initial criticality by the end of FY 1992. As for the demonstration fast breeder reactor (DFBR) of Japan, the Japan Atomic Power Company (JAPC) is promoting design study under the contracts with several leading Japanese fabricators, including Toshiba, Hitachi and Mitsubishi Heavy Industries, for selection of the basic specifications of DFBR. The related research and development (R and D) works are underway at several organizations under the discussion and coordination of the Japanese FBR R and D Steering Committee, which was established by the JAPAC, PNC, Japan Atomic Energy Research Institute (JAERI) and Central Research Institute of Electric Power Industry (CRIEPI). Progress of the design study and the related R and D are reported to the Subcommittee on FBR Development Program of JAEC. Recent major emphases on the PNC R and D are placed on the integrated feedback of all existing R and D results and experiences to the development of demonstration reactor. Furthermore, the overall functional and performance tests of Monju, is another important key role to attain further excellency of FBR technology, with

  11. Development Program of the Advanced HANARO Reactor in Korea

    International Nuclear Information System (INIS)

    Yang, I.-S.; Ahn, J.-H.; Han, K.-I.; Parh, C.; Jun, B.-J.; Kim, Y.-J.

    2006-01-01

    The development program of an advanced HANARO (AHR) reactor started in Korea to keep abreast of the increasing future demand, from both home and abroad, for research activities. This paper provides a review of the status of research reactors in Korea, the operating experience of the HANARO, the design principles and preliminary features of an advanced HANARO reactor, and the specific strategy of an advanced HANARO reactor development program. The design principles were established in order to design a new multi-purpose research reactor that is safe, economically competitive and technically feasible. These include the adaptation of the HANARO design concept, its operating experience, a high ratio of flux to power, a high degree of safety, improved economic efficiency, improved operability and maintainability, increased space and expandability, and ALARA design optimization. The strategy of an advanced HANARO reactor development program considers items such as providing a digital advanced HANARO reactor in cyber space, a method for the improving the design quality and economy of research reactors by using Computer Integrated Engineering, and more effective advertising using diverse virtual reality. This development program will be useful for promoting the understanding of and interest in the operating HANARO as well as an advanced HANARO reactor under development in Korea. It will provide very useful information to a country that may need a research reactor in the near future for the promotion of public health, bio-technology, drug design, pharmacology, material processing, and the development of new materials. (author)

  12. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1990-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

  13. Results and recommendations from the reactor chemistry and corrosion tasks of the reactor materials program

    International Nuclear Information System (INIS)

    Baumann, E.W.; Ondrejcin, R.S.

    1990-11-01

    Within the general context of extended service life, the Reactor Materials Program was initiated in 1984. This comprehensive program addressed material performance in SRS reactor tanks and the primary coolant or Process Water System (PWS) piping. Three of the eleven tasks concerned moderator quality and corrosion mitigation. Definition and control of the stainless steel aqueous environment is a key factor in corrosion mitigation. The Reactor Materials Program systematically investigated the SRS environment and its effect on crack initiation and propagation in stainless steel, with the objective of improving this environment. The purpose of this report is to summarize the contributions of Tasks 6, 7 and 10 of the Reactor Materials Program to the understanding and control of moderator quality and its relationship to mitigation of stress corrosion cracking

  14. Advanced test reactor testing experience-past, present and future

    International Nuclear Information System (INIS)

    Marshall, Frances M.

    2006-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

  15. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... comments were received. A companion guide, DG-1277, ``Initial Test Program of Emergency Core Cooling... NUCLEAR REGULATORY COMMISSION [NRC-2011-0129] Preoperational Testing of Emergency Core Cooling... (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

  16. Advanced reactor development: The LMR integral fast reactor program at Argonne

    International Nuclear Information System (INIS)

    Till, C.E.

    1990-01-01

    Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

  17. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1991-01-01

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  18. Plutonium recycle in PWR reactors (Brazilian Nuclear Program)

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-02-01

    An evaluation is made of the material requirements of the nuclear fuel cycle with plutonium recycle. It starts from the calculation of a reference reactor and allows the evaluation of demand under two alternatives of nuclear fuel cycle for Pressurized Water Reactors (PWR): without plutonium recycle; and with plutonium recycle. Calculations of the reference reactor have been carried out with the CELL-CORE codes. For plutonium recycle, the concept of uranium and plutonium homogeneous mixture has been adopted, using self-produced plutonium at equilibrium, in order to get minimum neutronic perturbations in the reactor core. The refueling model studied in the reference reactor was the 'out-in' scheme with a constant number of changed fuel elements (approximately 1/3 of the core). Variations in the material requirements were studied considering changes in the installed nuclear capacity of PWR reactors, the capacity factor of these reactors, and the introduction of fast breeders. Recycling plutonium produced inside the system can reach economies of about 5%U 3 O 8 and 6% separative work units if recycle is assumed only after the 5th operation cycle of the thermal reactors. The cumulative amount of fissile plutonium obtained by the Brazilian Nuclear Program of PWR reactors by 1991 should be sufficient for a fast breeder with the same capacity as Angra 2. For the proposed fast breeder programs, the fissile plutonium produced by thermal reactors is sufficient to supply fast breeder initial necessities. Howewer, U 3 O 8 and SWU economy with recycle is not significant when the proposed fast breeder program is considered. (Author) [pt

  19. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  20. Comparison of advanced reactors program of different international vendors

    International Nuclear Information System (INIS)

    Agnihotri, N.K.

    2001-01-01

    The full text follows. Proposal for presenting a paper on Advanced Reactor Program Given below is the abstract for Track 6 session on Advanced Reactor at the ninth International Conference on Nuclear Engineering being held in Nice, France from April 8. through 12. 2001. This paper will provide an update on Advanced Reactor Program of different vendors in the United States, Japan, and Europe. Specifically the paper will look at the history of different Advanced Reactor Programs, international experience, aspect of economy due to standardization, and the highlights of technical specifications. The paper will also review aspects of Economy due to standardization, public acceptance, required construction time, and the experience of different vendors. The objective of the presentation is to underscore the highlights of the Reactor Program of different vendors in order to keep the attendees of the conference up-to-date. The presentation will be an impartial overview from an outsider's (not part of the Nuclear Steam Supply System's staff). (author)

  1. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  2. DOE University Reactor Sharing Program. Renewal for 1994--1995

    International Nuclear Information System (INIS)

    Chappas, W.J.; Adams, V.G.

    1994-01-01

    The Department of Energy University Reactor Sharing Program at University of Maryland, College Park (UMCP) has, once again, stimulated a broad use of the reactor facilities by undergraduate and graduate students, visitors, and professionals. Participants are exposed to topics such as nuclear engineering, radiation safety, and nuclear reactor operations. This information is presented through various means including tours, slide presentations, experiments, and discussions. Student research using the MUTR is also encouraged. In addition, the Reactor Sharing Program here at the University of Maryland does not limit itself to the confines of the TRIGA reactor facility. Incorporated in the program are the Maryland University Neutron Activation Analysis Laboratory, the Maryland University Radiation Effects Laboratory, and the UMCP 2x4 Thermal Hydraulic Loop. These facilities enhance and give an added dimension to the tours and experiments

  3. Fuel rod-grid interaction wear: in-reactor tests (LWBR development program)

    International Nuclear Information System (INIS)

    Stackhouse, R.M.

    1979-11-01

    Wear of the Zircaloy cladding of LWBR irradiation test fuel rods, resulting from relative motion between rod and rod support contacts, is reported. Measured wear depths were small, 0.0 to 2.7 mils, but are important in fuel element behavior assessment because of the local loss of cladding thickness, as well as the effect on grid spring forces that laterally restrain the rods. An empirical wear analysis model, based on out-of-pile tests, is presented. The model was used to calculate the wear on the irradiation test fuel rods attributed to a combination of up-and-down motions resulting from power and pressure/temperature cycling of the test reactor, flow-induced vibrations, and assembly handling scratches. The calculated depths are generally deeper than the measured depths

  4. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  5. Fast Flux Test Facility: a unique resource for breeder reactor development

    International Nuclear Information System (INIS)

    Peckinpaugh, C.L.; Patterson, J.R.

    1984-08-01

    Based on encouraging data from the first two years of operation, the FFTF program has been redirected to close out testing on the intermediate lifetime fuel system and move directly to an early demonstration of a fuel system with a projected life of three to five years. Full advantage of favorable reactor physics effects will be taken to enhance inherent shutdown features and maximize reactor safety while reducing capital costs. Plant reliability will be demonstrated by achievement of high plant capacity factors

  6. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  7. Real time simulator for material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

    2012-03-15

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  8. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo

    2012-01-01

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  9. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    Conlin, J.A.

    1985-08-01

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

  10. US/DOE Man-Machine Integration program for liquid metal reactors

    International Nuclear Information System (INIS)

    D'Zmura, A.P.; Seeman, S.E.

    1985-03-01

    The United States Department of Energy (DOE) Man-Machine Integration program was started in 1980 as an addition to the existing Liquid Metal Fast Breeder Reactor safety base technology program. The overall goal of the DOE program is to enhance the operational safety of liquid metal reactors by optimum integration of humans and machines in the overall reactor plant system and by application of the principles of human-factors engineering to the design of equipment, subsystems, facilities, operational aids, procedures and environments. In the four years since its inception the program has concentrated on understanding the control process for Liquid Metal Reactors (LMRs) and on applying advanced computer concepts to this process. This paper describes the products that have been developed in this program, present computer-related programs, and plans for the future

  11. Advances by the Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs

  12. Advanced Instrumentation for Transient Reactor Testing

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

    2018-01-31

    Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

  13. A study on the development program of the advanced marine reactors

    International Nuclear Information System (INIS)

    Kobayashi, H.; Sako, K.; Iida, H.; Yamaji, A.

    1992-01-01

    JAERI has formulated two attractive concepts of advanced marine reactors. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 150 kWe DRX (Deep-sea Reactor X) for a deep sea research submersible. They adopt new technologies such as an integral type PWR, in-vessel type control rod drive mechanisms, a water-filled containment vessel and a passive decay heat removal system, which would enable to satisfy the essential requirements for marine reactors for next generation, i.e.; compact, light, highly passive safe and easy to operate. From now on, following conceptual design, the engineering design phase is going to start in order to advance the research and development of MRX and DRX further and to obtain the data necessary for the detail design and construction of the actual reactors. JAERI is studying on the program to develop the engineering design research on MRX and DRX, which consists mainly of the particularization of design, the data acquisition by experiments (synthetic hydrothermal dynamics experiments, fundamental tests related to passive core cooling and demonstration tests on reliability and operability), the development of particular components and the development of advanced design tools. (author)

  14. Proposal of world network on material testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Izumo, Hironobu; Hori, Naohiko; Ishitsuka, Etsuo; Ishihara, Masahiro

    2011-01-01

    Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

  15. Reactor safety issues resolved by the 2D/3D Program

    International Nuclear Information System (INIS)

    Damerell, P.S.; Simons, J.W.

    1993-07-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated

  16. NRC valve performance test program - check valve testing

    International Nuclear Information System (INIS)

    Jeanmougin, N.M.

    1987-01-01

    The Valve Performance Test Program addresses the current requirements for testing of pressure isolation valves (PIVs) in light water reactors. Leak rate monitoring is the current method used by operating commercial power plants to survey the condition of their PIVs. ETEC testing of three check valves (4-inch, 6-inch, and 12-inch nominal diameters) indicates that leak rate testing is not a reliable method for detecting impending valve failure. Acoustic emission monitoring of check valves shows promise as a method of detecting loosened internals damage. Future efforts will focus on evaluation of acoustic emission monitoring as a technique for determining check valve condition. Three gate valves also will be tested to evaluate whether the check valve results are applicable to gate type PIVs

  17. U.S. fast reactor materials and structures program

    International Nuclear Information System (INIS)

    Harms, W.O.; Purdy, C.M.

    1984-01-01

    The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program

  18. Quality assurance program plan for the Reactor Research Experiment Programs (RREP)

    International Nuclear Information System (INIS)

    Pipher, D.G.

    1982-05-01

    This document describes the Quality Assurance Program plans which will be applied to tasks on Reactor Research Experiments performed on Sandia National Laboratories' reactors. The program provides for individual project or experiment quality plan development and allows for reasonable plan flexibility and maximum plan visibility. Various controls and requirements in this program plan are considered mandatory on all features which are identified as important to public health and safety (Level I). It is the intent of this document that the Quality Assurance program comprise those elements which will provide adequate assurance that all components, equipment, and systems of the experiments will perform as designed, and hence prevent delays and costs due to rejections or failures

  19. US Department of Energy 1992--1993 Reactor Sharing Program

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1994-04-01

    The University of Florida Training Reactor serves as a host institution to support various educational institutions which are located primarily within the state of Florida. All users and uses were carefully screened to assure the usage was for educational institutions eligible for participation in the Reactor Sharing Program. Three tables are included that provide basic information about the 1992--1993 program and utilization of the reactor facilities by user institutions

  20. Light water reactor pressure isolation valve performance testing

    International Nuclear Information System (INIS)

    Neely, H.H.; Jeanmougin, N.M.; Corugedo, J.J.

    1990-07-01

    The Light Water Reactor Valve Performance Testing Program was initiated by the NRC to evaluate leakage as an indication of valve condition, provide input to Section XI of the ASME Code, evaluate emission monitoring for condition and degradation and in-service inspection techniques. Six typical check and gate valves were purchased for testing at typical plant conditions (550F at 2250 psig) for an assumed number of cycles for a 40-year plant lifetime. Tests revealed that there were variances between the test results and the present statement of the Code; however, the testing was not conclusive. The life cycle tests showed that high tech acoustic emission can be utilized to trend small leaks, that specific motor signature measurement on gate valves can trend and indicate potential failure, and that in-service inspection techniques for check valves was shown to be both feasible and an excellent preventive maintenance indicator. Life cycle testing performed here did not cause large valve leakage typical of some plant operation. Other testing is required to fully understand the implication of these results and the required program to fully implement them. (author)

  1. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  2. JENDL-3.3 thermal reactor benchmark test

    International Nuclear Information System (INIS)

    Akie, Hiroshi

    2001-01-01

    Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

  3. Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

    International Nuclear Information System (INIS)

    Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

    2004-01-01

    Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

  4. A fast-running fuel management program for a CANDU reactor

    International Nuclear Information System (INIS)

    Choi, Hangbok

    2000-01-01

    A fast-running fuel management program for a CANDU reactor has been developed. The basic principle of this program is to select refueling channels such that the reference reactor conditions are maintained by applying several constraints and criteria when selecting refueling channels. The constraints used in this program are the channel and bundle power and the fuel burnup. The final selection of the refueling channel is determined based on the priority of candidate channels, which enhances the reactor power distribution close to the time-average model. The refueling simulation was performed for a natural uranium CANDU reactor and the results were satisfactory

  5. OSU Reactor Sharing Program FY 1995 annual report

    International Nuclear Information System (INIS)

    Higginbotham, J.F.

    1996-10-01

    This is the annual report of the activities supported under the Oregon State University Reactor Sharing Program, award number DE-FG06-NE38137. The beginning date for the award was September, 30, 1995 and the end date was September 29, 1996. Work conducted under this award is internally administered at the Radiation Center through a project tasking system. This allows for excellent quality control for the work which is performed from the point of initial contact, through the reactor application, project report generation and financial accounting. For the current fiscal year, FY95, the total cost of the reactor sharing program, including Radiation Center contributions, was $66,323.20 of which $40,000.00 was supplied by the DOE Reactor Sharing Program. The details of individual project costs is given in Table 1. The work performed for the individual projects are described in the brief work descriptions given in Table 2

  6. Development status and potential program for development of proliferation-resistant molten-salt reactors

    International Nuclear Information System (INIS)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review

  7. Development status and potential program for development of proliferation-resistant molten-salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review.

  8. Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-15

    A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

  9. University Reactor Matching Grants Program

    International Nuclear Information System (INIS)

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-01-01

    During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given

  10. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  11. Feasibility study of the Dragon reactor for HTGR fuel testing

    International Nuclear Information System (INIS)

    Wallroth, C.F.

    1975-01-01

    The Organization of European Community Development (OECD) Dragon high-temperature reactor project has performed HTGR fuel and fuel element testing for about 10 years. To date, a total of about 250 fuel elements have been irradiated and the test program continues. The feasibility of using this test facility for HTGR fuel testing, giving special consideration to U. S. needs, is evaluated. A detailed description for design, preparation, and data acquisition of a test experiment is given together with all possible options on supporting work, which could be carried out by the experienced Dragon project staff. 11 references. (U.S.)

  12. A continuing success - The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Mustin, Tracy P.; Clapper, Maureen; Reilly, Jill E.

    2000-01-01

    The United States Department of Energy, in consultation with the Department of State, adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program, established under this policy, has completed 16 spent fuel shipments. 2,651 material test reactor (MTR) assemblies, one Slowpoke core containing less than 1 kilogram of U.S.-origin enriched uranium, 824 Training, Research, Isotope, General Atomic (TRIGA) rods, and 267 TRIGA pins from research reactors around the world have been shipped to the United States so far under this program. As the FRR SNF Acceptance Program progresses into the fifth year of implementation, a second U.S. cross country shipment has been completed, as well as a second overland truck shipment from Canada. Both the cross country shipment and the Canadian shipment were safely and successfully completed, increasing our knowledge and experience in these types of shipments. In addition, two other shipments were completed since last year's RERTR meeting. Other program activities since the last meeting included: taking pre-emptive steps to avoid license amendment pitfalls/showstoppers for spent fuel casks, publication of a revision to the Record of Decision allowing up to 16 casks per ocean going vessel, and the issuance of a cable to 16 of the 41 eligible countries reminding their governments and the reactor operators that the U.S.-origin uranium in their research reactors may be eligible for return to the United States under the Acceptance Program and urging them to begin discussions on shipping schedules. The FRR SNF program has also supported the Department's implementation of the competitive pricing policy for uranium and resumption of shipments of fresh uranium for fabrication into assemblies for research reactors. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues

  13. Program of RA reactor start-up to nominal power; Program dizanja reaktora 'RA' na nominalnu snagu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is desed in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members.

  14. DOE University Reactor Sharing Program. Final technical report for 1996--1997

    International Nuclear Information System (INIS)

    Chappas, W.J.; Adams, V.G.

    1998-01-01

    The Department of Energy University Reactor Sharing Program at University of Maryland, College Park (UMCP) has, once again, stimulated a broad use of the reactor and radiation facilities by undergraduate and graduate students, visitors, and professionals. Participants are exposed to topics such as nuclear engineering, radiation safety, and nuclear reactor operations. This information is presented through various means including tours, slide presentations, experiments, and discussions. Student research using the MUTR is also encouraged. In addition, the Reactor Sharing Program here at the University of Maryland does not limit itself to the confines of the TRIGA reactor facility. Incorporated in the program are the Maryland University Radiation Effects Laboratory, and the UMCP 2 x 4 Thermal Hydraulic Loop. These facilities enhance and give an added dimension to the tours and experiments. The Maryland University Training Reactor (MUTR) and the associated laboratories are made available to any interested institution six days a week on a scheduled basis. Most institutions are scheduled at the time of their first request--a reflection of their commitment to the Reactor Sharing Program. The success of the past years by no means guarantees future success. Therefore, the reactor staff is more aggressively pursuing its outreach program, especially with junior colleges and universities without reactor or radiation facilities; more aggressively developing demonstration and training programs for students interested in careers in nuclear power and radiation technology; and more aggressively up-grading the reactor facilities--not only to provide a better training facility but to prepare for relicensing in the year 2000

  15. An overview of the Indian program related to fast reactor core mechanical behaviour

    International Nuclear Information System (INIS)

    Govindarajan, S.; Bhoje, S.B.; Paranjpe, S.R.

    1984-01-01

    This Indian review paper presents the evolution of the fast breeder program which began with fast breeder test reactor (FBTR) commencing in 1972. The state-of-art in the field of core mechanical behaviour is reviewed

  16. The advanced test reactor national scientific user facility advancing nuclear technology

    International Nuclear Information System (INIS)

    Allen, T.R.; Thelen, M.C.; Meyer, M.K.; Marshall, F.M.; Foster, J.; Benson, J.B.

    2009-01-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  17. The UMR reactor outreach program for expanded educational utilization

    International Nuclear Information System (INIS)

    Freeman, D.; Bolon, A.

    1992-01-01

    In recent years, the University of Missouri-Rolla Reactor (UMRR) facility has been under intense financial scrutiny by the university administration; primarily due to ever-tightening budgets and declines in nuclear engineering (NE) enrollment. In response to criticisms of low utilization, the reactor staff has developed and implemented a dynamic outreach program designed to significantly increase the educational role of the facility on campus. The outreach program is based on the principle that the potential to provide service to the UMR community is far in excess of the present level of service. The program is designed to identify and inform potential users of how their courses or programs can be augmented through use of the reactor facility. The net effect of the outreach program is greater campus communication and awareness of the unique capabilities as applied to each discipline. A natural product of the outreach program should be increased research

  18. Computer programs for the in-core fuel management of power reactors

    International Nuclear Information System (INIS)

    1981-08-01

    This document gives a survey of the presently tested and used computer programs applicable to the in-core fuel management of light and heavy water moderated nuclear power reactors. Each computer program is described (provided that enough information was supplied) such that the nature of the physical problem solved and the basic mathematical or calculational approach are evident. In addition, further information regarding computer requirements, up-to-date applications and experiences and specific details concerning implementation, staff requirements, etc., are provided. Program procurement conditions, possible program implementation assistance and commercial conditions (where applicable) are given. (author)

  19. Seismic isolation development for the US advanced liquid-metal reactor program

    International Nuclear Information System (INIS)

    Gluekler, E.L.; Bigelow, C.C.; DeVita, V.; Kelly, J.M.; Seidensticker, R.W.; Tajirian, F.F.

    1989-01-01

    GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US Department of Energy (DOE). The objective of this development is to provide, by the turn of the century, a reactor concept with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactor and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long duration earthquakes. Seismic isolation is accomplished with high damping natural rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity. This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program is described in this paper and selected results are presented. The initial testing indicated excellent performance of high damping natural rubber bearings

  20. Participation in the U.S. Department of Energy Reactor Sharing Program

    International Nuclear Information System (INIS)

    Mulder, R. U.; Benneche, P. E.; Hosticka, B.

    1998-01-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these users institutions is enhanced by the use of the nuclear facilities

  1. Japanese contributions to ITER testing program of solid breeder blankets for DEMO

    International Nuclear Information System (INIS)

    Kuroda, Toshimasa; Yoshida, Hiroshi; Takatsu, Hideyuki; Maki, Koichi; Mori, Seiji; Kobayashi, Takeshi; Suzuki, Tatsushi; Hirata, Shingo; Miura, Hidenori.

    1991-04-01

    ITER Conceptual Design Activity (CDA), which has been conducted by four parties (Japan, EC, USA and USSR) since May 1988, has been finished on December 1990 with a great achievement of international design work of the integrated fusion experimental reactor. Numerous issues of physics and technology have been clarified for providing a framework of the next phase of ITER (Engineering Design Activity; EDA). Establishment of an ITER testing program, which includes technical test issues of neutronics, solid breeder blankets, liquid breeder blankets, plasma facing components, and materials, has been one of the goals of the CDA. This report describes Japanese proposal for the testing program of DEMO/power reactor blanket development. For two concepts of solid breeder blanket (helium-cooled and water-cooled), identification of technical issues, scheduling of test program, and conceptual design of test modules including required test facility such as cooling and tritium recovery systems have been carried out as the Japanese contribution to the CDA. (author)

  2. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  3. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  4. Reliability test for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Uchiyama, Junichi

    1998-01-01

    41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

  5. Space reactor fuel element testing in upgraded TREAT

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  6. Ground testing of an SP-100 prototypic reactor

    International Nuclear Information System (INIS)

    Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

    1988-01-01

    SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

  7. Space reactor fuel element testing in upgraded TREAT

    Science.gov (United States)

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

    1993-01-01

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

  8. Non-linear programming method in optimization of fast reactors

    International Nuclear Information System (INIS)

    Pavelesku, M.; Dumitresku, Kh.; Adam, S.

    1975-01-01

    Application of the non-linear programming methods on optimization of nuclear materials distribution in fast reactor is discussed. The programming task composition is made on the basis of the reactor calculation dependent on the fuel distribution strategy. As an illustration of this method application the solution of simple example is given. Solution of the non-linear program is done on the basis of the numerical method SUMT. (I.T.)

  9. Fabrication and testing of uranium nitride fuel for space power reactors

    Science.gov (United States)

    Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

    1988-02-01

    Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

  10. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  11. Status and program of development of the fast breeder reactor system in the U.S

    International Nuclear Information System (INIS)

    Roberts, R.W.

    1977-01-01

    The U.S. Energy Research and Development Administration's highest priority breeder concept is the Liquid Metal Fast Breeder Reactor (LMFBR). Its objective is the development of a broad technological and engineering base with extensive utility and industrial involvement which will establish a timely capability for a competitive and commercial breeder industry. In addition to technological development, an integral part of the LMFBR program is the transfer of LMFBR technology to the nuclear industry and the parallel development of utility capability. An important indicator of the effectiveness of this program is the successful construction and operation of a number of LMFBR experimental and test reactors and associated non-nuclear test facilities such as those located at Santa Susana, California, the Argonne (III). National Laboratory and the Hanford Engineering Development Laboratory at Richland, Washington. A principal element of the U.S. fast breeder reactor program is the Fast Flux Test Facility (FFTF). The largest test reactor facility in the world, the FFTF will focus on the testing of fuels and material. The FFTF, being built at the Hanford Engineering Development Laboratory, is scheduled for completion in the late 1970's. The next step toward timely commercial breeder capability is to implement a program for a large-scale demonstration of this concept. A 380 megawatt electrical demonstration plant - designated the Clinch River Breeder Reactor Plant (CRBRP) - will be constructed and operated near Oak Ridge, Tennessee, under a cooperative arrangement of industrial contractors, utilities, and the Government. The completion date is 1983. The estimated cost of the CRBRP project is 1.95 billion dollars. This includes the cost of design, construction, related research and development, and five years of operation - the full timespan from the project's beginning in 1972 through completion. The Nation's principal electric utilities have strongly endorsed the

  12. Design of A Vibration and Stress Measurement System for an Advanced Power Reactor 1400 Reactor Vessel Internals Comprehensive Vibration Assessment Program

    International Nuclear Information System (INIS)

    Ko, Doyoung; Kim, Kyuhyung

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea

  13. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    Directory of Open Access Journals (Sweden)

    DO-YOUNG KO

    2013-04-01

    Full Text Available In accordance with the US Nuclear Regulatory Commission (US NRC, Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP has been developed for an Advanced Power Reactor 1400 (APR1400. The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment. Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

  14. Experience in reactor research and development programs as educational system for thermohydraulic engineering

    International Nuclear Information System (INIS)

    Zaki, G.M.; Fikry, M.M.

    1977-01-01

    A reactor development program within a research reactor facility can be used for personnel training on the operation of power reactors and research in the different fields of nuclear science and engineering. A training program is proposed where reactor maintenance and operation, in addition to conducting development programs and executing projects, are utilized for forming specialized groups. The paper gives a short survey of a heat transfer program where out of pile and in-core studies are conducted along with two-phase flow investigations. This program covers the main requirements for WWR (water cooled and moderated reactor) power uprating and furnishes basic knowledge on power reactor thermal parameters. The major facilities for conducting similar programs devoted to education are mentioned

  15. Nuclear Reactor Sharing Program

    International Nuclear Information System (INIS)

    1994-01-01

    The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

  16. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

    2004-01-01

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  17. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  18. Inspection program for U.S. research reactors

    International Nuclear Information System (INIS)

    Isaac, Patrick J.

    2010-01-01

    This paper presents an established program for inspection of nuclear research reactors to ensure that systems and techniques are in accordance with regulatory requirements and to provide protection for the health and safety of the public. The inspection program, implemented from the time a facility gets licensed, remains in effect through operations, shutdown, decommissioning, and until the license is terminated. The program establishes inspection methodology for operating, safeguards, and decommissioning activities. Using a performance- based approach, inspectors focus their attention on activities important to safety. Inspection procedures allow the inspectors to assess facility safety and compliance to applicable requirements. A well designed inspection program is an integral part of the mechanism to ensure that the level of performance in the strategic areas of reactor safety, radiation safety, and safeguards is acceptable and provides adequate protection of public health and safety. (author)

  19. Integral Fast Reactor Program

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  20. Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

    The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

  1. Trends in light water reactor dosimetry programs

    International Nuclear Information System (INIS)

    Rahn, F.J.; Serpan, C.Z.; Fabry, A.; McElroy, W.N.; Grundl, J.A.; Debrue, J.

    1977-01-01

    Dosimetry programs and techniques play an essential role in the continued assurance of the safety and reliability of components of light water reactors. Primary concern focuses on the neutron irradiation embrittlement of reactor pressure vessels and methods by which the integrity of a pressure vessel can be predicted and monitored throughout its service life. Research in these areas requires a closely coordinated program which integrates the elements of the calculational and material sciences, the development of advanced dosimetric techniques and the use of benchmarks and validation of these methods. The paper reviews the status of the various international efforts in the dosimetry area

  2. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies

  3. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.

    2011-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program's strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  4. Preliminary Options Assessment of Versatile Irradiation Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

  5. Proceedings of the international symposium on materials testing reactors

    International Nuclear Information System (INIS)

    Ishihara, Masahiro; Kawamura, Hiroshi

    2009-01-01

    This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

  6. Opening address at the international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    Kazuo Sato

    1984-01-01

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of variuos programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. More detailed status of the RERTR program in Japan, as the host country is covered in this presentation

  7. The experimental program of neutronphysics for advanced water reactors

    International Nuclear Information System (INIS)

    Martin-Deider, L.; Cathalu, S.; Santamarina, A.; Gomit, M.

    1985-11-01

    The C.E.A. and E.D.F. has jointly undertaken a program of experimental studies on under-moderated water lattices, with mixed oxide fuel UO 2 -PuO 2 . Undermoderated lattices offer high conversion ratios. This type of lattice could limit in the future the natural uranium consumption of pressurized water reactors. This experimental program is aimed at qualifying neutron transport calculations in a large range of moderating ratio (between 0.5 and 1.5). It includes three experiments: ERASME, a critical experiment of large size in the EOLE reactor at Cadarache; ICARE, an irradiation experiment in the MELUSINE reactor at Grenoble; and an experiment to measure the reactivity effects by oscillations in the MINERVE reactor at Cadarache [fr

  8. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses.

  9. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    International Nuclear Information System (INIS)

    1985-07-01

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses

  10. 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)

  11. Reactor safety issues resolved by the 2D/3D program

    International Nuclear Information System (INIS)

    1995-09-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author)

  12. Reactor safety issues resolved by the 2D/3D program

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author).

  13. Proceedings of the international meeting on development, fabrication, and application of Reduced Enrichment fuels for Research and Test Reactors (RERTR). Base technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-08-01

    The international effort to develop new fuel materials and designs which will make it feasible to fuel research and test reactors throughout the world with low-enrichment uranium, instead of high-enrichment uranium, has made significant progress during the past year. This progress has taken place at research centers located in many different countries, and is of crucial interest to reactor operators and licensors whose geographical distribution is even more varied. It is appropriate, therefore, that international meetings be held periodically to foster direct communication among the specialists in this area. To achieve this purpose, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the third of a series which begun in 1978. The papers presented at this meeting were divided into sessions according to relevant subject: status of RERTR program and safety issues; development of new fuel types; testing of new fuel elements; specific reactor applications. These proceedings were edited by various members of the RERTR Program.

  14. Proceedings of the international meeting on development, fabrication, and application of Reduced Enrichment fuels for Research and Test Reactors (RERTR). Base technology

    International Nuclear Information System (INIS)

    1983-08-01

    The international effort to develop new fuel materials and designs which will make it feasible to fuel research and test reactors throughout the world with low-enrichment uranium, instead of high-enrichment uranium, has made significant progress during the past year. This progress has taken place at research centers located in many different countries, and is of crucial interest to reactor operators and licensors whose geographical distribution is even more varied. It is appropriate, therefore, that international meetings be held periodically to foster direct communication among the specialists in this area. To achieve this purpose, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the third of a series which begun in 1978. The papers presented at this meeting were divided into sessions according to relevant subject: status of RERTR program and safety issues; development of new fuel types; testing of new fuel elements; specific reactor applications. These proceedings were edited by various members of the RERTR Program

  15. CRIEPI test program for seismic isolation of the FBR

    International Nuclear Information System (INIS)

    Shiojiri, Hiroo

    1989-01-01

    This paper describes the Central Research Institute of Electric Power Industry's (CRIEPIs) seismic isolation program. The test and research program on seismic isolation was started in 1987 by CRIEPI under contract with the Ministry of International Trade and Industry (MITI) of Japan. It was intended to establish a technical basis for the application of seismic isolation to fast breeder reactors (FBRs). In this paper, some details of the program and results of the preliminary study are described

  16. 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.

  17. 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

  18. Nuclear reactor safety program in US department of energy and future perspectives

    International Nuclear Information System (INIS)

    Song, Y.T.

    1988-01-01

    The US Department of Energy (DOE) establishes policy, issues orders, and assures compliance with requirements. The contractors who design, construct, modify, operate, maintain and decommission DOE reactors, set forth the assessment of the safety of cognizant reactors and implement DOE orders. Teams of experts in the Department, through scheduled and unscheduled review programs, reassess the safety of reactors in every phases of their lives. As new technology develops, the safety programs are reevaluated and policies are modified to accommodate these new technologies. The diagnostic capabilities of the computer using multiple alarms to enhance detection of defects and control of a reactor have been greatly utilized in reactor operating systems. The Application of artificial intelligence technologies for diagnostic and even for the decision making process in the event of reactor accidents would be one of the future trends in reactor safety programs

  19. TEMP-M program for thermal-hydraulic calculation of fast reactor fuel assemblies

    International Nuclear Information System (INIS)

    Bogoslovskaya, C.P.; Sorokin, A.P.; Tikhomirov, B.B.; Titov, P.A.; Ushakov, P.A.

    1983-01-01

    TEMP-M program (Fortran, BESM-6 computer) for thermal-hydraulic calculation of fast reactor fuel assemblies is described. Results of calculation of temperature field in a 127 fuel element assembly of BN-600, reactor accomplished according to TEMP-N program are considered as an example. Algorithm, realized in the program, enables to calculate the distributions of coolant heating, fuel element temperature (over perimeter and length) and assembly shell temperature. The distribution of coolant heating in assembly channels is determined from a solution of the balance equation system which accounts for interchannel exchange, nonadiabatic conditions on the assembly shell. The TEMP-M program gives necessary information for calculation of strength, seviceability of fast reactor core elements, serves an effective instrument for calculations when projecting reactor cores and analyzing thermal-hydraulic characteristics of operating reactor fuel assemblies

  20. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    1983-08-01

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  1. Licensing experience of the HTR-10 test reactor

    International Nuclear Information System (INIS)

    Sun, Y.; Xu, Y.

    1996-01-01

    A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

  2. Status of Wrought FeCrAl-UO2 Capsules Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harp, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Core, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction, and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.

  3. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

    1999-01-01

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

  4. The BR2 materials testing reactor. Past, ongoing and under-study upgradings

    Energy Technology Data Exchange (ETDEWEB)

    Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

    1990-05-01

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

  5. PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

    International Nuclear Information System (INIS)

    Leon, D.M.; White, M.D.; Moore, P.A.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of this report is to make the reduced instrument responses during test 104 available

  6. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, January--March 1976

    Energy Technology Data Exchange (ETDEWEB)

    Zane, J. O.; Farman, R. F.; Hanson, D. J.; Peterson, A. C.; Ybarrondo, L. J.; Berta, V. T.; Naff, S. A.; Crocker, J. G.; Martinson, Z. R.; Smolik, G. R.; Cawood, G. W.; Quapp, W. J.; Ramsthaler, J. H.; Ransom, V. H.; Scofield, M. P.; Dearien, J. A.; Bohn, M. P.; Burnham, B. W.; James, S. W.; Lee, W. H.; Lime, J. F.; Nalezny, C. L.; MacDonald, P. E.; Thompson, L. B.; Domenico, W. F.; Rice, R. E.; Hendrix, C. E.; Davis, C. B.

    1976-06-01

    Light water reactor sfaety research performed January through March 1976 is summarized. Results of the Semiscale Mod-1 blowdown heat transfer test series relating to those phenomena that influence core fluid and heat transfer effects are analyzed, and preliminary analyses of the recently completed reflood heat transfer test series are summarized for the forced and gravity feed reflood tests. The first nonnuclear LOCE in the LOFT program was successfully completed and preliminary results are presented. Preliminary results are given for the PCM 8-1 RF Test, the PCM-2A Test, and the Irradiation Effects Scoping Test 2 in the Thermal Fuel Behavior Program. Model development and verification efforts reported in the Reactor Behavior Program include checkout of RELAP4/MOD5 Update 1, development of a new hydrodynamic model for two-phase separated flows, development of the RACHET code to assess the assumptions in current fuel behavior codes of uniform stress and strain in the cladding, modifications of the containment code BEACON, analysis of results from the Halden Assembly IFA-429 helium sorption experiment, development of correlations for the thermal conductivity of UO/sub 2/ and (U,Pu)O/sub 2/, and evaluation of RALAP4 through comparison of calculated results with data from the GE Blowdown Heat Transfer and Semiscale experiments.

  7. Seismic isolation development for the US advanced liquid-metal reactor program

    International Nuclear Information System (INIS)

    Gluekler, E.L.; Bigelow, C.C.; DeVita, V.; Kelly, J.M.; Seidensticker, R.W.; Tajirian, F.F.

    1991-01-01

    GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US DOE. The objective of this development is to provide, by the turn of the century, a reactor with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactors and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long-duration earthquakes. Seismic isolation is accomplished with high-damping natural-rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity. This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program, which is supported by the US DOE, ANL, Energy Technology Engineering Center (ETEC), the University of California at Berkeley (UC-Berkeley), GE, and Bechtel National, Inc. (BNI), is described and selected results are presented. The initial testing indicated excellent performance of high-damping natural-rubber bearings. The development of seismic isolation guidelines is in progress as a joint activity between ENEA of Italy and the GE Team. (orig./HP)

  8. Participation in the US Department of Energy Reactor Sharing Program

    International Nuclear Information System (INIS)

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report

  9. TRIGLAV-W a Windows computer program package with graphical users interface for TRIGA reactor core management calculations

    International Nuclear Information System (INIS)

    Zagar, T.; Zefran, B.; Slavic, S.; Snoj, L.; Ravnik, M.

    2006-01-01

    TRIGLAV-W is a program package for reactor calculations of TRIGA Mark II research reactor cores. This program package runs under Microsoft Windows operating system and has new friendly graphical user interface (GUI). The main part of the package is the TRIGLAV code based on two dimensional diffusion approximation for flux distribution calculation. The new GUI helps the user to prepare the input files, runs the main code and displays the output files. TRIGLAV-W has a user friendly GUI also for the visualisation of the calculation results. Calculation results can be visualised using 2D and 3D coloured graphs for easy presentations and analysis. In the paper the many options of the new GUI are presented along with the results of extensive testing of the program. The results of the TRIGLAV-W program package were compared with the results of WIMS-D and MCNP code for calculations of TRIGA benchmark. TRIGLAV-W program was also tested using several libraries developed under IAEA WIMS-D Library Update Project. Additional literature and application form for TRIGLAV-W program package beta testing can be found at http://www.rcp.ijs.si/triglav/. (author)

  10. Requalification of SPERT [Special Power Excursion Reactor Test] pins for use in university reactors

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Domagala, R.F.; Dates, L.R.

    1986-12-01

    A series of nondestructive and destructive examinations have been performed on a representative sample of stainless steel-clad UO 2 fuel pins procured in the early-to-mid 1960s for the SPERT program. These examinations were undertaken in order to requalify the SPERT pins for use in converting university research reactors from the use of highly enriched uranium to the use of low-enriched uranium. The requalification program included visual and dimensional inspections of fuel pins and fuel pellets, radiographic inspections of welds, fill gas analyses, and chemical and spectrographic analyses of fuel and cladding materials. In general all attributes tested were within or very close to specified values, although some weld defects not covered by the original specifications were found. 1 ref., 4 figs., 11 tabs

  11. Program for analyzing power boost tests

    International Nuclear Information System (INIS)

    Wills, C.A.

    1982-03-01

    A rapid increase of power in a reactor produces a failure in the fuel. Experiments to study the conditions in the NRU reactor after such failures have been planned and carried out. Given the concentrations of specified isotopes at a number of times over the length of an experiment as produced for example, from the program SARGS and the power history of the reactor, this program calculates the release rates, escape rate coefficients, and fractional releases for the isotopes. These values may be optionally printed and plotted. Decay schemes for a limited number of mass numbers are implemented. The program is written in FORTRAN and runs on the CDC 6600 - CYBER 170 system

  12. Cover gas seals: FFTF-LMFBR seal test program

    International Nuclear Information System (INIS)

    Kurzeka, W.; Oliva, R.; Welch, T.S.; Shimazaki, T.

    1974-01-01

    The objectives of this program are to: (1) conduct static and dynamic tests to demonstrate or determine the mechanical performance of full-size (cross section) FFTF fuel transfer machine and reactor vessel head seals intended for use in a sodium vapor-inert gas environment, (2) demonstrate that these FFTF seals or new seal configurations provide acceptable fission product and cover gas retention capabilities at Clinch River Breeder Reactor Plant (CRBRP) operating environmental conditions other than radiation, and (3) develop improved seals and seal technology for the CRBRP to support the national objective to reduce all atmospheric contaminations to low levels

  13. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  14. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately.

  15. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately

  16. Procedures and techniques for the management of experimental fuels from research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1999-04-01

    Almost all countries that have undertaken fuel development programs for power, research or military reactors have experimental and exotic fuels, either stored at the original research reactors where they have been tested or at some away-from-reactor storage facility. These spent fuel liabilities cannot follow the standard treatment recognized for modern power reactor fuels. They include experimental and exotic fuels ranging from liquids to coated spheres and in configurations ranging from full test assemblies to post irradiation examination specimens set in resin. This document contains an overview of the extent of the problem of managing experimental and exotic fuels from research and test reactors and an expert evaluation of the overall situation in countries which participated in the meeting

  17. Acoustic emission monitoring of preservice testing at Watts Bar Unit 1 Nuclear Reactor

    International Nuclear Information System (INIS)

    Hutton, P.H.; Pappas, R.A.; Friesel, M.A.

    1985-02-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Plant in the US during hot functional preservice testing is described. Background, methodology, and results are included. The work discussed here is a major milestone in a program supported by the US NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing to AE monitoring during reactor operation. 3 refs., 6 figs

  18. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Katoh, Yutai [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Wilson, Dane F [ORNL

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of

  19. An Idea of 20% test of the Initial Core Reactor Physics

    International Nuclear Information System (INIS)

    Roh, Kyung Ho; Yang, Sung Tae; Jung, Ji Eun

    2012-01-01

    Many tests have been performed on the OPR1000 and APR1400 before commercial operation. Some of these tests were performed at reactor power levels of 20% and 50%. The CPC (Core Protection Calculator) power distribution test is one of these tests. It is performed to assure the reliability of the Core Protection Calculator System (CPCS). Through this test, SAM1 is calculated using the snapshots2. The test takes about nine hours at a reactor power level of 20% and about thirty hours at a reactor power level of 50%. SAM used at each reactor power level is as follows: 1. Reactor power of 0% ∼ 20%: Designed SAM (DSAM) 2. Reactor power of 20% ∼ 50%: SAM calculated (C-SAM) at a reactor power of 20% 3. Reactor power 50% ∼ End of Cycle : SAM calculated at a reactor power of 50% As mentioned earlier, SAM is calculated and punched into CPC to assure the reliability of CPCS. Therefore, CPC is operated having penalties with D-SAM until3 reaching a reactor power of 20%. That is, the penalty of CPC will be removed when SAM is calculated and punched into the CPC at a reactor power of 20%. But these penalties are considered to be removed after a reactor power of 50% test in order to maintain the conservatism of the CPC. This is done because the final values calculated using C-SAM, in contrast to those calculated using SAM, a reactor power of 50%, are not correct. This paper began from an idea, 'If so, what would happen if we removed the CPC power distribution test at a reactor power of 20%?'

  20. Reactor operator screening test experiences

    International Nuclear Information System (INIS)

    O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

    1976-01-01

    When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

  1. Emergency medical assistance programs for nuclear power reactors

    International Nuclear Information System (INIS)

    Linnemann, R.E.; Mettler, F.A. Jr.

    1977-01-01

    This paper deals with a simple but practical medical support of geographically distributed nuclear reactors in isolated areas. A staff of experts at a centre devote their full attention to accident prevention and preparedness at reactor sites. They establish and maintain emergency medical programs at reactor sites and nearby support hospitals. The emphasis is on first aid and emergency treatment by medical attendants who are not and cannot be experts in radiation but do know how to treat patients. (author)

  2. 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

  3. University Reactor Sharing Program

    International Nuclear Information System (INIS)

    Reese, W.D.

    2004-01-01

    Research projects supported by the program include items such as dating geological material and producing high current super conducting magnets. The funding continues to give small colleges and universities the valuable opportunity to use the NSC for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy. The Reactor Sharing Program has supported the construction of a Fast Neutron Flux Irradiator for users at New Mexico Institute of Mining and Technology and the University of Houston. This device has been characterized and has been found to have near optimum neutron fluxes for A39/Ar 40 dating. Institution final reports and publications resulting from the use of these funds are on file at the Nuclear Science Center

  4. Environmental Fatigue of Metallic Materials in Nuclear Power Plants - A Review of Korean Test Programs

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Changheul; Jang, Hun; Hong, Jongdae [KAIST, Daejeon (Korea, Republic of); Cho, Hyunchul [Doosan Heavy Industry and Construction, Changwon (Korea, Republic of); Kim, Tae Soon; Lee, Jaegon [KHNP, Daejeon (Korea, Republic of)

    2013-12-15

    Environmental fatigue of the metallic components in light water reactors has been the subject of extensive research and regulatory interest in Korea and abroad. Especially, it was one of the key domestic issues for the license renewal of operating reactors and licensing of advanced reactors during the early 2000s. To deal with the environmental fatigue issue domestically, a systematic test program has been initiated and is still underway. The materials tested were SA508 Gr.1a low alloy steels, 316LN stainless steels, cast stainless steels, and an Alloy 690 and 52M weld. Through tests and subsequent analysis, the mechanisms of reduced low cycle fatigue life have been investigated for those alloys. In addition, the effects of temperature, dissolved oxygen level, and dissolved hydrogen level on low cycle fatigue behaviors have been investigated. In this paper, the test results and key analysis results are briefly summarized. Finally, an on-going test program for hot-bending of 347 stainless steel is introduced.

  5. Testing of seismic isolation bearings for advanced liquid metal reactor prism

    International Nuclear Information System (INIS)

    Tajirian, F.F.; Kelly, J.M.

    1988-01-01

    Seismic isolation can significantly mitigate earthquake loads on liquid metal reactors (LMR), thus reducing the impact of seismic loads on design. This improves plant safety margins for beyond-design basis seismic events and enhances adaptability of a standardized design to a variety of sites, with potential cost benefits. The PRISM (Power Reactor Inherently Safe Module) LMR incorporates a horizontal isolation system which consists of high damping steel laminated rubber bearings. The results of an experimental program to determine the mechanical properties of the rubber compound and the bearing performance under different loading conditions are presented. The test results demonstrate the excellent performance of the bearings and their suitability for isolating compact LMR plants

  6. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

    2000-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  7. Reactor group constants and benchmark test

    Energy Technology Data Exchange (ETDEWEB)

    Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

  8. Data on test results of vessel cooling system of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2003-02-01

    High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

  9. Tests of vacuum interrupters for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Warren, R.; Parsons, M.; Honig, E.; Lindsay, J.

    1979-04-01

    The Tokamak Fusion Test Reactor (TFTR) project at Princeton University requires the insertion of a resistor in an excited ohmic-heating coil circuit to produce a plasma initiation pulse (PIP). It is expected that the maximum duty for the switching system will be an interruption of 24 kA with an associated recovery voltage of 25 kV. Vacuum interrupters were selected as the most economical means to satisfy these requirements. However, it was felt that some testing of available systems should be performed to determine their reliability under these conditions. Two interrupter systems were tested for over 1000 interruptions each at 24 kA and 25 kV. One system employed special Westinghouse type WL-33552 interrupters in a circuit designed by LASL. This circuit used a commercially available actuator and a minimum size counterpulse bank and saturable reactor. The other used Toshiba type VGB2-D20 interrupters actuated by a Toshiba mechanism in a Toshiba circuit using a larger counterpulse bank and saturable reactor

  10. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

    1996-09-01

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

  11. LMFBR. Off normal, transient test facilities and programs in the USA

    International Nuclear Information System (INIS)

    Herbst, R.J.

    1985-01-01

    The United States fast breeder reactor development program has included operational transient analyses and experiments to verify the predicted performance of core components. Operational transient testing has focused on off-normal operation during Plant Protection System terminated transient-overpower events. In-pile and out-of-pile tests have been used to simulate predicted thermal and mechanical strain cycles and measure component response. The spectrum of reactivity ramp rates investigated in TOP tests has recently been expanded to include rates of less than $0.1/s. These slow ramp rate studies are being done in cooperation with the Japanese. The US has also cooperated with the UK in the transient testing of Prototype Fast Reactor fuel pins

  12. Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

    International Nuclear Information System (INIS)

    Atkins, D.F.; Lillie, A.F.

    1983-01-01

    As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

  13. Russian research reactor fuel return program starts shipping fuel to Russia

    International Nuclear Information System (INIS)

    Dedik, T.; Bolshinsky, I.; Krass, A.

    2003-01-01

    For almost four years the United States (U.S), the Russian Federation (R.F.), and the International Atomic Energy Agency (IAEA) have been discussing an initiative to return Soviet/Russian-origin research reactor fuel to the Russian Federation. In a series of bilateral and trilateral meetings in Vienna and Moscow, considerable progress has been made toward defining the Russian Research Reactor Fuel Return Program as well as obtaining the necessary technical data to facilitate the return. More than 20 research reactors in 17 countries that have Soviet- or Russian-supplied fuel have identified. Most of these reactors have stocks of both fresh and irradiated HEU fuel that must be carefully stored and managed for many years to come. On September 21, 2003 the Russian Research Reactor Fuel Return program shipped 14 kg of fresh Russian-origin HEU fuel from Romania to the nuclear fuel fabrication facility in Russia, which represented the beginning of the practical implementation of the program. (author)

  14. Sandia Pulse Reactor-IV Project

    International Nuclear Information System (INIS)

    Reuscher, J.A.

    1983-01-01

    Sandia National Laboratories has developed, designed and operated fast burst reactors for over 20 years. These reactors have been used for a variety of radiation effects programs. During this period, programs have required larger irradiation volumes primarily to expose complex electronic systems to postulated threat environments. As experiment volumes increased, a new reactor was built so that these components could be tested. The Sandia Pulse Reactor-IV is a logical evolution of the two decades of fast burst reactor development at Sandia

  15. Conceptual design for simulator of irradiation test reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Ohto, Tsutomu; Magome, Hirokatsu; Izumo, Hironobu; Hori, Naohiko

    2012-03-01

    A simulator of irradiation test reactors has been developed since JFY 2010 for understanding reactor behavior and for upskilling in order to utilize a nuclear human resource development (HRD) and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR, one of the irradiation test reactors, and it simulates operation, irradiation tests and various kinds of accidents caused by the reactor and irradiation facility. The development of the simulator is sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. The training using the simulator will be started for the nuclear HRD from JFY 2012. This report summarizes the result of the conceptual design of the simulator in JFY 2010. (author)

  16. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  17. Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

    International Nuclear Information System (INIS)

    Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

    1981-01-01

    Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

  18. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  19. Gas cooled fast reactor control rod drive mechanism deceleration unit. Test program

    International Nuclear Information System (INIS)

    Wagner, T.H.

    1981-10-01

    This report presents the results of the airtesting portion of the proof-of-principle testing of a Control Rod Scram Deceleration Device developed for use in the Gas Cooled Fast Reactor (GCFR). The device utilizes a grooved flywheel to decelerate the translating assembly (T/A). Two cam followers on the translating assembly travel in the flywheel grooves and transfer the energy of the T/A to the flywheel. The grooves in the flywheel are straight for most of the flywheel length. Near the bottom of the T/A stroke the grooves are spiraled in a decreasing slope helix so that the cam followers accelerate the flywheel as they transfer the energy of the falling T/A. To expedite proof-of-principle testing, some of the materials used in the fabrication of certain test article components were not prototypic. With these exceptions the concept appears to be acceptable. The initial test of 300 scrams was completed with only one failure and the failure was that of a non-prototypic cam follower outer sleeve material

  20. Development of a Test Facility to Simulate the Reactor Flow Distribution of APR+

    International Nuclear Information System (INIS)

    Euh, D. J.; Cho, S.; Youn, Y. J.; Kim, J. T.; Kang, H. S.; Kwon, T. S.

    2011-01-01

    Recently a design of new reactor, APR+, is being developed, as an advanced type of APR1400. In order to analyze the thermal margin and hydraulic characteristics of APR+, quantification tests for flow and pressure distribution with a conservation of flow geometry are necessary. Hetsroni (1967) proposed four principal parameters for a hydraulic model representing a nuclear reactor prototype: geometry, relative roughness, Reynolds number, and Euler number. He concluded that the Euler number should be similar in the prototype and model under the preservation of the aspect ratio on the flow path. The effect of the Reynolds number at its higher values on the Euler number is rather small, since the dependency of the form and frictional loss coefficients on the Reynolds number is seen to be small. ABB-CE has carried out several reactor flow model test programs, mostly for its prototype reactors. A series of tests were conducted using a 3/16 scale reactor model. (see Lee et al., 2001). Lee et al (1991) performed experimental studies using a 1/5.03 scale reactor flow model of Yonggwang nuclear units 3 and 4. They showed that the measured data met the acceptance criteria and were suitable for their intended use in terms of performance and safety analyses. The design of current test facility was based on the conservation of Euler number which is a ratio of pressure drop to dynamic pressure with a sufficiently turbulent region having a high Reynolds number. By referring to the previous study, the APR+ design is linearly reduced to 1/5 ratio with a 1/2 of the velocity scale, which yields a 1/39.7 of Reynolds number scaling ratio. In the present study, the design feature of the facilities, named 'ACOP', in order to investigate flow and pressure distribution are described

  1. United States Department of Energy commercial reactor spent fuel programs being conducted at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Piscitella, R.R.; Rasmussen, T.L.; Uhl, D.L.

    1987-01-01

    The Idaho National Engineering Laboratory participation in OCRWM programs includes the Spent Fuel Storage Cask Testing Program, Dry Rod Consolidation Technology Program, Prototypical Consolidation Demonstration Program, the Nuclear Fuel Services Project, and the Cask Systems Acquisition Program. The DOE has entered into a cooperative agreement with Virginia Power and the Electric Power Research Institute to demonstrate storage of commercial spent fuel in steel storage casks. The Program conducted heat transfer and shielding tests with three storage casks with intact spent fuel assemblies and two casks with consolidated spent fuel rods, one of which was previously tested with intact fuel, and provides test information in support of Virginia Power's at-reactor dry storage licensing effort. 3 figs., 1 tab

  2. Reliability tests for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

    1998-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  3. The safety basis of the integral fast reactor program

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Seidel, B.R.

    1990-01-01

    The Integral Fast Reactor (IFR) and metallic fuel have emerged as the US Department of Energy reference reactor concept and fuel system for the development of an advanced liquid-metal reactor. This article addresses the basic elements of the IFR reactor concept and focuses on the safety advances achieved by the IFR Program in the areas of (1) fuel performance, (2) superior local faults tolerance, (3) transient fuel performance, (4) fuel-failure mechanisms, (5) performance in anticipated transients without scram, (6) core-melt mitigation, and (7) actinide recycle

  4. Method and program for complex calculation of heterogeneous reactor

    International Nuclear Information System (INIS)

    Kalashnikov, A.G.; Glebov, A.P.; Elovskaya, L.F.; Kuznetsova, L.I.

    1988-01-01

    An algorithm and the GITA program for complex one-dimensional calculation of a heterogeneous reactor which permits to conduct calculations for the reactor and its cell simultaneously using the same algorithm are described. Multigroup macrocross sections for reactor zones in the thermal energy range are determined according to the technique for calculating a cell with complicate structure and then the continuous multi group calculation of the reactor in the thermal energy range and in the range of neutron thermalization is made. The kinetic equation is solved using the Pi- and DSn- approximations [fr

  5. FISS: a computer program for reactor systems studies

    International Nuclear Information System (INIS)

    Tamm, H.; Sherman, G.R.; Wright, J.H.; Nieman, R.E.

    1979-08-01

    ΣFISSΣ is a computer code for use in investigating alternative fuel cycle strategies for Canadian and world nuclear programs. The code performs a system simulation accounting for dynamic effects of growing nuclear systems. Facilities in the model include storage for irradiated fuel, mines, plants for enrichment, fuel fabrication, fuel reprocessing and heavy water, and reactors. FISS is particularly useful for comparing various reactor strategies and studying sensitivities of resource consumption, capital investment and energy costs with changes in fuel cycle parameters, reactor parameters and financial variables. (author)

  6. GCRA review and appraisal of HTGR reactor-core-design program

    International Nuclear Information System (INIS)

    1980-09-01

    The reactor-core-design program has as its principal objective and responsibility the design and resolution of major technical issues for the reactor core and core components on a schedule consistent with the plant licensing and construction program. The task covered in this review includes three major design areas: core physics, core thermal and hydraulic performance fuel element design, and in-core fuel performance evaluation

  7. Experiences in stability testing of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Otaduy, P.J.

    1986-01-01

    The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  9. Breeder reactor program in the USA

    International Nuclear Information System (INIS)

    Brewer, S.

    1978-01-01

    In the United States, commercial fuel reprocessing and demonstration test of plutonium breeder reactors were now postponed. LMFBR project and schedule of FFTF and afterwards await the results of INFCE. However, this is not discarding the development of LMFBRs. With the existing energy resources, the United States can have the opportunity to make careful and thorough study. Emphasis is placed on the research and development on new alternative types of fuel. FFTF going to be operated soon should provide effective means for the developments of FBR fuel and materials. High priority is to be retained for the test and development of sodium system hardwares. The nuclear proliferation problem is not related to heat transfer and secondary systems; it is associated with the selection of fuel and fuel cycle. The whole program is centered around LMFBR design and development. The target output will be 600 x 10 3 -- 700 x 10 3 MW. In the United States, now is the time to develop excellent products and to study the nuclear proliferation problem more carefully. (Mori, K.)

  10. High-temperature gas-cooled reactor safety-reliability program plan

    Energy Technology Data Exchange (ETDEWEB)

    1981-03-01

    The purpose of this document is to present a safety plan as part of an overall program plan for the design and development of the High Temperature Gas-Cooled Reactor (HTGR). This plan is intended to establish a logical framework for identifying the technology necessary to demonstrate that the requisite degree of public risk safety can be achieved economically. This plan provides a coherent system safety approach together with goals and success criterion as part of a unifying strategy for licensing a lead reactor plant in the near term. It is intended to provide guidance to program participants involved in producing a technology base for the HTGR that is fully responsive to safety consideration in the design, evaluation, licensing, public acceptance, and economic optimization of reactor systems.

  11. In-core fuel management programs for nuclear power reactors

    International Nuclear Information System (INIS)

    1984-10-01

    In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

  12. Fusion Reactor Safety Research Program annual report, FY-79

    International Nuclear Information System (INIS)

    Crocker, J.G.; Cohen, S.

    1980-08-01

    The objective of the program is the development, coordination, and execution of activities related to magnetic fusion devices and reactors that will: (a) identify and evaluate potential hazards, (b) assess and disclose potential environmental impacts, and (c) develop design standards and criteria that eliminate, mitigate, or reduce those hazards and impacts. The program will provide a sound basis for licensing fusion reactors. Included in this report are portions of four reports from two outside contractors, discussions of the several areas in which EG and G Idaho is conducting research activities, a discussion of proposed program plan development, mention of special tasks, a review of fusion technology program coordination by EG and G with other laboratories, and a brief view of proposed FY-80 activities

  13. Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

    2007-03-01

    The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

  14. The European fusion program and the role of the research reactors

    International Nuclear Information System (INIS)

    Laesser, R.; Andreani, R.; Diegele, E.

    2005-01-01

    The main objectives of the European long-term Fusion Technology Program are i) investigation of DEMO breeding blankets options, ii) development of low activation materials resistant to high neutron fluence, iii) construction of IFMIF for validation of DEMO materials, and iv) promotion of modelling efforts for the understanding of radiation damage. A large effort is required for the development and performance verification of the materials subjected to the intense neutron irradiation encountered in fusion reactors. In the absence of a strong 14.1 MeV neutron source fission materials research reactors are used. Elaborate in-pile and post-irradiation examinations are performed. In addition, the modelling effort is increased to predict the damage by a 'true' fusion spectrum in the future. Even assuming that a positive decision for IFMIF construction can be reached, the operation of a limited number of materials test reactors is needed to perform irradiation studies on large samples and for screening. (author)

  15. Summary of ROSA-4 LSTF first phase test program and station blackout (TMLB) test results

    International Nuclear Information System (INIS)

    Tasaka, K.; Kukita, Y.; Anoda, Y.

    1990-01-01

    This paper summarizes major test results obtained at the ROSA-4 Large Scale Test Facility (LSTF) during the first phase of the test program. The results from a station blackout (TMLB) test conducted at the end of the first-phase program are described in some detail. The LSTF is an integral test facility being operated by the Japan Atomic Energy Research Institute for simulation of pressurized water reactor (PWR) thermal-hydraulic responses during small-break loss-of-coolant accidents (SBLOCAs) and operational/abnormal transients. It is a 1/48 volumetrically scaled, full-height, full-pressure simulator of a Westinghouse-type 4-loop PWR. The facility includes two symmetric primary loops each one containing an active inverted-U tube steam generator and an active reactor coolant pump. The loop horizontal legs are sized to conserve the scaled (1/24) volumes as well as the length to the square root of the diameter ratio in order to simulate the two-phase flow regime transitions. The primary objective of the LSTF first-phase program was to define the fundamental PWR thermal-hydraulic responses during SBLOCAs and transients. Most of the tests were conducted with simulated component/operator failures, including unavailability of the high pressure injection system and auxiliary feedwater system, as well as operator failure to take corrective actions. The forty-two first phase tests included twenty-nine SBLOCA tests conducted mainly for cold leg breaks, three abnormal transient tests and ten natural circulation tests. Attempts were made in several of the SBLOCA tests to simulate the plant recovery procedures as well as candidate accident management measures for prevention of high-pressure core melt situation. The natural circulation tests simulated the single-phase and two-phase natural circulation as well as reflux condensation behavior in the primary loops in steady or quasi-steady states

  16. Material and geometry options and performance characteristics for a test reactor

    International Nuclear Information System (INIS)

    Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

    1993-01-01

    For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

  17. Program for tests on magnetic bearing suspended rotor dynamics for gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Takada, Shoji; Takizuka, Takakazu; Kunitomi, Kazuhiko; Kosugiyama, Shinichi; Yan, Xing

    2003-01-01

    A program for test on rotor dynamics was planned for the turbo-machine of the Gas Turbine High Temperature Reactor (GTHTR300). The rotor system of the turbo-machine consists of a turbo-compressor rotor and a generator rotor connected with a flexible coupling, each suspended with two radial magnetic bearings. The rotors, which are flexible rotors, pass over the critical speeds of bending mode. The magnetic bearing is required to have a high load capacity, about 10 times larger than any built thus far to support a flexible rotor. In the rotor design, the standard limit of the vibration amplitude of 75 μm at the rated rotational speed of 3,600 rpm was fulfilled by optimizing the stiffness of the magnetic bearings. A test apparatus was designed to verify the design of the magnetic bearing suspended turbo-machine rotor of the GTHTR300. The test apparatus is composed of 1/3-scale test rotors, which are connected with a flexible coupling and driven by a variable speed motor. The test magnetic bearing was designed within the state-of-the-art technology to have a load capacity about 1/10 of that of the actual one. The test rotors were designed to closely simulate the critical speeds and vibration modes of the actual ones. This paper shows the test apparatus and the test plan for the magnetic bearing suspended rotor system. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  18. Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

    International Nuclear Information System (INIS)

    Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

    2004-01-01

    Full text: The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by (1) implementing safety upgrades, or (2) assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

  19. Cost effective safety enhancements for research reactors in Uzbekistan and Kazakhstan - results of a joint program with US DOE

    International Nuclear Information System (INIS)

    Earle, O.K.; Carlson, R.B.; Rakhmanov, A.; Salikhbaev, U.S.; Chernyaev, V.; Chakrov, P.

    2004-01-01

    The US Department of Energy's Office of International Nuclear Safety and Cooperation established the Integrated Research Reactor Safety Enhancement Program (IRRSEP) in February 2002 to support U.S. nonproliferation goals by implementing safety upgrades, or assisting with the safe shutdown and decommissioning of foreign test and research reactors which present security concerns. IRRSEP's key program components are: Phase I: Self-evaluation by facility using provided checklists followed by prioritization to identify the 20 highest risk facilities; Phase II: Site visits with technical evaluation to finalize a list of projects that will enhance safety consistent with IAEA observations; Phase III: Corrective measures to implement the projects. Phases I, II and III are accomplished on a rolling basis, such that work is ongoing at three or four reactors per year. IRRSEP's key objective is to resolve the highest-priority nuclear safety issues at the most vulnerable foreign research reactors as quickly as possible. The prioritization methodology employed identified which research reactors fell into this category. The corrective measures mutually developed with the host facility are based on the premise of developing a sustainable infrastructure within each country to deal with its own nuclear material safety, security, and response issues in the future. IRRSEP also assists in creating an international framework of cooperation and openness between research and test reactor operators, and national and international regulators. The initial projects under IRRSEP are underway at research reactors in Kazakhstan, Uzbekistan, and Romania. This paper focuses on the projects undertaken at the WWR-K research reactor at the Institute of Nuclear Physics in Alatau, Kazakhstan and the WWR-SM research reactor at the Institute of Nuclear Physics in Ulugbek, Uzbekistan. These projects demonstrate the success and cost effectiveness of the IRRSEP program

  20. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  1. Nuclear reactor safety program in U.S. Department of Energy and future perspectives

    International Nuclear Information System (INIS)

    Song, Y.T.

    1987-01-01

    The U.S. Department of Energy (DOE) establishes policy, issues orders, and assures compliance with requirements. The contractors who design, construct, modify, operate, maintain and decommission DOE reactors, set forth the assessment of the safety of cognizant reactors and impliment DOE orders. Teams of experts in the Depatment, through scheduled and unscheduled review programs, reassess the safety of reactors in every phases of their lives. As new technology develops, the safety programs are reevaluated and policies are modified to accommodate these new technologies. The diagnostic capabilities of the computer using multiple alarms to enhance detection of defects and control of a reactor have been greatly utilized in reactor operating systems. The application of artificial intelligence (AI) technologies for diagnostic and even for the decision making process in the event of reactor accidents would be one of the future trends in reactor safety programs. (author)

  2. Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

    International Nuclear Information System (INIS)

    Whitman, G.D.

    1975-01-01

    The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references

  3. Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program. PSF Blind Test workshop minutes. Summary

    International Nuclear Information System (INIS)

    Guthrie, G.L.; Lippincott, E.P.; McGarry, E.D.

    1984-01-01

    A ''Blind Test'' workshop was held on April 9-11, 1984, at the Holiday Inn in Richland, WA. At the workshop, participant groups compared ''Blind'' calculations with existing data which was unavailable to them at the time the calculations were made. The purpose of the exercise was to allow each participant group to test the group's ability to predict ''in-wall'' mechanical property degradation for a simulated nuclear reactor pressure vessel irradiation

  4. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    International Nuclear Information System (INIS)

    MH Lane

    2006-01-01

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations

  5. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  6. Methodological efficiency of reactor pressure vessel control samples programs and reliability of WWER vessels at the Ukraine NPP

    International Nuclear Information System (INIS)

    Kovbasenko, S.N.; Brednev, V.I.; Zinchenko, O.Ya.

    2000-01-01

    An analysis is made for the programs with the use of check specimens (CS) for reactor vessel (RV) control which is considered as the only method for estimating embrittlement in RV material and its operational capability. Advantages and disadvantages of the programs are noted. Other methods are developed and improved to estimate properties of RV metal and welded joints. It is shown that the most valid information about RV metal state can be obtained only by testing specimens prepared from templets cut directly from a RV wall in the region of reactor core. Ukrainian researchers designed the equipment and performed successful tests on templet cutting from the surface of the metal to study mechanical properties of critical pipelines [ru

  7. Users guide to the computer program FURST (FUture Reactor STrategies)

    International Nuclear Information System (INIS)

    Hatton, H.

    1981-01-01

    A program has been written to calculate the future resource requirements for the nuclear portion of an electricity generating system. Starting from a given total energy demand projection the program calculates the required growth of the electrical generating system, the total nuclear system, and the portion provided by reactors using an advanced fuel cycle by successive application of the Fisher/Pry penetration formula. Several options are available. These include the ability to (1) change the growth rates of any part of the system; (2) change the characteristics of the reactors; (3) include the effects of decommissioning reactors at the end of their design lifetimes; (4) vary the date of introduction of advanced reactors; and (5) limit the amount of natural uranium available annually. The output gives the history of the growth of the nuclear system and the uranium mining and fuel reprocessing requirements. The output can be obtained either as tables of numbers or graphs with crossplots to compare reactor systems or total energy scenarios. (author)

  8. Implosion and staging systems for a Scyllac Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gribble, R.F.; Linford, R.K.; Thomassen, K.I.

    1976-01-01

    The implosion heating and adiabatic compression processes will be separated in future theta pinch devices. The circuit to achieve the fast implosion heating and power crowbar (staging) for the Scyllac Fusion Test Reactor is described here. The plasma is very tightly coupled to the circuit and presents a varying inductive load. Computer-aided circuit designs which achieve a programmed magnetic field waveform are described. The field approximates a two-step waveform, on-off-on, which is ideal for achieving the large initial plasma radius needed for stability. The components for the circuits have been developed and are being tested in experiments at Los Alamos

  9. Implosion and staging systems for a Scyllac fusion test reactor

    International Nuclear Information System (INIS)

    Gribble, R.F.; Linford, R.K.; Thomassen, K.I.

    1975-01-01

    The implosion heating and adiabatic compression processes will be separated in future theta pinch devices. The circuit to achieve the fast implosion heating and power crowbar (staging) for the Scyllac Fusion Test Reactor is described here. The plasma is very tightly coupled to the circuit and presents a varying inductive load. Computer-aided circuit designs which achieve a programmed magnetic field waveform are described. The field approximates a two-step waveform, on-off-on, which is ideal for achieving the large initial plasma radius needed for stability. The components for the circuits have been developed and are being tested in experiments at Los Alamos. (auth)

  10. The Jules Horowitz reactor, a new high performance European material testing reactor open to international users: present status and objectives

    International Nuclear Information System (INIS)

    Iracane, D.; Bignan, G.

    2010-01-01

    The development of nuclear power as a sustainable and competitive energy source will continue to require research and development of fuel and material behaviour under irradiation. This necessitates a high performance material testing reactor (MTR). Facing the obsolescence of most of the existing MTR in Europe, France decided a few years ago the construction of the RJH (Jules Horowitz reactor). RJH is designed, built and will be operated as an international user facility. A first set of experimental hosting devices is being designed. For instance, there are the in-core CALIPSO Nak integrated loop for material studies and other loops for fuel studies under nominal or off-normal or accidental conditions. The RJH international program will focus on the following subjects: -) fuel reliability, assessed through power ramps tests and post-irradiation examination; -) Loss of coolant tests done out-of-pile in a first phase and in-pile in a possible second phase; and -) source term tests addressing fission products release. The paper reports also the point of view of VATTENFALL (a Swedish power utility), as a potential European RJH user. (A.C.)

  11. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    Moore, Mark; George, Robert; Spence, Harry; Nguyen, John

    1992-01-01

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  12. Programming for a nuclear reactor instrument simulation

    International Nuclear Information System (INIS)

    Cohn, C.

    1988-01-01

    This note discusses 8086/8087 machine-language programming for simulation of nuclear reactor instrument current inputs by means of a digital-analog converter (DAC) feeding a bank of series input resistors. It also shows FORTRAN programming for generating the parameter tales used in the simulation. These techniques would be generally useful for high-speed simulation of quantities varying over many orders of magnitude

  13. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  14. FFTF reactor assembly system technology

    International Nuclear Information System (INIS)

    Mangelsdorf, T.A.

    1975-01-01

    An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs

  15. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  16. Mechanical properties test program on structural materials in a sodium environment

    International Nuclear Information System (INIS)

    Natesan, K.; Chopra, O.K.; Kassner, T.F.

    1979-10-01

    This document describes in detail the ongoing and planned US Test program on the mechanical properties of sodium-exposed Type 316 austenitic stainless and Fe-2 1/4 Cr-1 Mo ferritic steels. The test program is based on the Development Requirement Specifications (DRS) established by the DOE/Clinch River Breeder Reactor Project (CRBRP) Program Office, the general need for the development of LMFBR structural-design criteria established by the Nuclear Systems Materials Handbook, and the need for a fundamental understanding of materials behavior in a sodium environment, which is generic to LMFBR systems. The planned test program is an extension of work based on current knowledge of sodium chemistry and the influence of sodium purity on the mechanical properties of structural materials

  17. Automated reactor protection testing saves time and avoids errors

    International Nuclear Information System (INIS)

    Raimondo, E.

    1990-01-01

    When the Pressurized Water Reactor units in the French 900MWe series were designed, the instrumentation and control systems were equipped for manual periodic testing. Manual reactor protection system testing has since been successfully replaced by an automatic system, which is also applicable to other instrumentation testing. A study on the complete automation of process instrumentation testing has been carried out. (author)

  18. Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

    International Nuclear Information System (INIS)

    Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

    1980-01-01

    Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

  19. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  20. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    International Nuclear Information System (INIS)

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-01-01

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  1. EPRI program in water reactor safety

    International Nuclear Information System (INIS)

    Loewenstein, W.B.; Gelhaus, F.; Gopalakrishnan, A.

    1975-01-01

    The basis for EPRI's water reactor safety program is twofold. First is compilation and development of fundamental background data necessary for quantified light-water reactor (LWR) safety assurance appraisals. Second is development of realistic and experimentally bench-marked analytical procedures. The results are expected either to confirm the safety margins in current operating parameters, and to identify overly conservative controls, or, in some cases, to provide a basis for improvements to further minimize uncertainties in expected performance. Achievement of these objectives requires the synthesis of related current and projected experimental-analytical projects toward establishment of an experimentally-based analysis for the assurance of safety for LWRs

  2. IAEA activities related to research reactor fuel conversion and spent fuel return programs

    International Nuclear Information System (INIS)

    Goldman, Ira N.; Adelfang, Pablo; Ritchie, Iain G.

    2005-01-01

    The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programs have supported research reactor fuel conversion from HEU to low enriched uranium (LEU), and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. (author)

  3. Present status and future perspective of research and test reactors in JAERI

    International Nuclear Information System (INIS)

    Baba, Osamu; Kaieda, Keisuke

    1999-01-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  4. Present status and future perspective of research and test reactors in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

    1999-08-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  5. Status of University of Cincinnati reactor-site nuclear engineering graduate programs

    International Nuclear Information System (INIS)

    Anno, J.N.; Christenson, J.M.; Eckart, L.E.

    1993-01-01

    The University of Cincinnati (UC) nuclear engineering program faculty has now had 12 yr of experience in delivering reactor-site educational programs to nuclear power plant technical personnel. Currently, with the sponsorship of the Toledo-Edison Company (TED), we are conducting a multiyear on-site graduate program with more than 30 participants at the Davis-Besse nuclear power plant. The program enables TED employees with the proper academic background to earn a master of science (MS) degree in nuclear engineering (mechanical engineering option). This paper presents a brief history of tile evolution of UC reactor-site educational programs together with a description of the progress of the current program

  6. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

  7. Optimization programs for reactor core fuel loading exhibiting reduced neutron leakage

    International Nuclear Information System (INIS)

    Darilek, P.

    1991-01-01

    The program MAXIM was developed for the optimization of the fuel loading of WWER-440 reactors. It enables the reactor core reactivity to be maximized by modifying the arrangement of the fuel assemblies. The procedure is divided into three steps. The first step includes the passage from the three-dimensional model of the reactor core to the two-dimensional model. In the second step, the solution to the problem is sought assuming that the multiplying properties, or the reactivity in the zones of the core, vary continuously. In the third step, parameters of actual fuel assemblies are inserted in the ''continuous'' solution obtained. Combined with the program PROPAL for a detailed refinement of the loading, the program MAXIM forms a basis for the development of programs for the optimization of fuel loading with burnable poisons. (Z.M.). 16 refs

  8. BWR Full Integral Simulation Test (FIST) program: facility description report

    International Nuclear Information System (INIS)

    Stephens, A.G.

    1984-09-01

    A new boiling water reactor safety test facility (FIST, Full Integral Simulation Test) is described. It will be used to investigate small breaks and operational transients and to tie results from such tests to earlier large-break test results determined in the TLTA. The new facility's full height and prototypical components constitute a major scaling improvement over earlier test facilities. A heated feedwater system, permitting steady-state operation, and a large increase in the number of measurements are other significant improvements. The program background is outlined and program objectives defined. The design basis is presented together with a detailed, complete description of the facility and measurements to be made. An extensive component scaling analysis and prediction of performance are presented

  9. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1, 1977--September 30, 1977

    International Nuclear Information System (INIS)

    1977-01-01

    Work covered includes an updated listing of the alloys selected for the screening tests, plus complete test specimen matrices for the screening program. The present design and construction status of the simulated reactor helium loops and testing and analysis facilities and equipment are discussed. Also covered are the loading matrices for the screening creep tests

  10. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Energy Technology Data Exchange (ETDEWEB)

    Korobkina, E., E-mail: ekorobk@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Medlin, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Wehring, B.; Hawari, A.I. [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Huffman, P.R.; Young, A.R. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Beaumont, B. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Palmquist, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States)

    2014-12-11

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ∼40K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  11. Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor

    International Nuclear Information System (INIS)

    Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon

    2005-01-01

    A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams

  12. Utilization of fission reactors for fusion engineering testing

    International Nuclear Information System (INIS)

    Deis, G.A.; Miller, L.G.

    1985-01-01

    Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful

  13. Design of high temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

    1994-09-01

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

  14. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    International Nuclear Information System (INIS)

    Agarwal, Vivek; Smith, James A.; Jewell, James Keith

    2015-01-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  15. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  16. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection

  17. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  18. Tokamak engineering test reactor

    International Nuclear Information System (INIS)

    Conn, R.W.; Jassby, D.L.

    1975-07-01

    The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

  19. Imperfection detection probability at ultrasonic testing of reactor vessels

    International Nuclear Information System (INIS)

    Kazinczy, F. de; Koernvik, L.Aa.

    1980-02-01

    The report is a lecture given at a symposium organized by the Swedish nuclear power inspectorate on February 1980. Equipments, calibration and testing procedures are reported. The estimation of defect detection probability for ultrasonic tests and the reliability of literature data are discussed. Practical testing of reactor vessels and welded joints are described. Swedish test procedures are compared with other countries. Series of test data for welded joints of the OKG-2 reactor are presented. Future recommendations for testing procedures are made. (GBn)

  20. Present status of Japan materials testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  1. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi

    2012-01-01

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  2. 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

  3. Light Water Reactor Sustainability Program Integrated Program Plan

    International Nuclear Information System (INIS)

    Griffith, George; Youngblood, Robert; Busby, Jeremy; Hallbert, Bruce; Barnard, Cathy; McCarthy, Kathryn

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  4. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  5. Reactor cover gas monitoring at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Bechtold, R.A.; Holt, F.E.; Meadows, G.E.; Schenter, R.E.

    1986-09-01

    The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the US Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100-day operating cycle began in April 1982 and the eighth operating cycle was completed in July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification

  6. Advanced Gas Cooled Reactor Materials Program. Reducing helium impurity depletion in HTGR materials testing

    International Nuclear Information System (INIS)

    Baldwin, D.H.

    1984-08-01

    Moisture depletion in HTGR materials testing rigs has been empirically studied in the GE High Temperature Reactor Materials Testing Laboratory (HTRMTL). Tests have shown that increased helium flow rates and reduction in reactive (oxidizable) surface area are effective means of reducing depletion. Further, a portion of the depletion has been shown to be due to the presence of free C released by the dissociation of CH 4 . This depletion component can be reduced by reducing the helium residence time (increasing the helium flow rate) or by reducing the CH 4 concentration in the test gas. Equipment modifications to reduce depletion have been developed, tested, and in most cases implemented in the HTRMTL to date. These include increasing the Helium Loop No. 1 pumping capacity, conversion of metallic retorts and radiation shields to alumina, isolation of thermocouple probes from the test gas by alumina thermowells, and substitution of non-reactive Mo-TZM for reactive metallic structural components

  7. Analytical evaluation on loss of off-side electric power simulation of the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Tachibana, Yukio; Takada, Eiji; Kunitomi, Kazuhiko

    2000-03-01

    A rise-to-power test of the high temperature engineering test reactor (HTTR) started on September 28 in 1999 for establishing and upgrading the technological basis for the high temperature gas-cooled reactor (HTGR). A loss of off-site electric power test of the HTTR from the normal operation under 15 and 30 MW thermal power will be carried out in the rise-to-power test. Analytical evaluations on transient behaviors of the reactor and plant during the loss of off-site electric power were conducted. These estimations are proposed as benchmark problems for the IAEA coordinated research program on 'Evaluation of HTGR Performance'. This report describes an event scenario of transient during the loss of off-site electric power, the outline of major components and system, detailed thermal and nuclear data set for these problems and pre-estimation results of the benchmark problems by an analytical code 'ACCORD' for incore and plant dynamics of the HTGR. (author)

  8. the JHR Material Testing Reactor

    International Nuclear Information System (INIS)

    Roure, C.; Cornu, B.; Berthet, B.; Simon, E.; Estre, N.; Guimbal, P.; Kinnunen, P.; Kotiluoto, P.

    2013-06-01

    The Jules Horowitz Reactor (JHR) is a European experimental reactor under construction in CEA Cadarache. It will be dedicated to material and fuel irradiation tests, and to medical isotopes production. Non-Destructive nuclear Examinations systems (NDE) will be implemented in pools to analyse the irradiated fuel or tested material in their supporting experimental irradiation devices extracted from the core or its immediate periphery. The Nuclear Measurement Laboratory (NML) of CEA Cadarache is working in collaboration with VTT (Technical Research Centre in Finland) in designing and developing NDE systems implementing gamma-ray spectroscopy and high energy X-ray imaging of the sample and irradiation device. CEA is also designing a neutron radiography system for which NML is working on the detection system. Design studies are performed with Monte Carlo transport codes and specific simulation tools developed by the NML for Xray and neutron imaging. (authors)

  9. 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

  10. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    International Nuclear Information System (INIS)

    Abe, K.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    1998-01-01

    A Japan-USA program of irradiation experiments for fusion research, ''JUPITER'', has been established as a 6 year program from 1995 to 2000. The goal is to study ''the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment''. This is phase-three of the collaborative program, which follows RTNS-II program (phase-1: 1982-1986) and FFTF/MOTA program (phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA program. JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. (orig.)

  11. Safety program considerations for space nuclear reactor systems

    International Nuclear Information System (INIS)

    Cropp, L.O.

    1984-08-01

    This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given

  12. Multifrequency tests in the EBR-II reactor plant

    International Nuclear Information System (INIS)

    Feldman, E.E.; Mohr, D.; Gross, K.C.

    1989-01-01

    A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

  13. Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Freedman, J.M.; Humphreys, J.R.

    1975-10-01

    A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

  14. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  15. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  16. Visual interface for the automation of the instrumented pendulum of Charpy tests used in the surveillance program of reactors vessel of nuclear power plants

    International Nuclear Information System (INIS)

    Rojas S, A.S.; Sainz M, E.; Ruiz E, J.A.

    2004-01-01

    Inside the Programs of Surveillance of the nuclear power stations periodic information is required on the state that keep the materials with those that builds the vessel of the reactor. This information is obtained through some samples or test tubes that are introduced inside the core of the reactor and it is observed if its physical characteristics remain after having been subjected to the radiation changes and temperature. The rehearsal with the instrumented Charpy pendulum offers information on the behavior of fracture dynamics of a material. In the National Institute of Nuclear Research (ININ) it has an instrumented Charpy pendulum. The operation of this instrument is manual, having inconveniences to carry out rehearsals with radioactive material, handling of high and low temperatures, to fulfill the normative ones for the realization of the rehearsals, etc. In this work the development of a computational program is presented (virtual instrument), for the automation of the instrumented pendulum. The system has modules like: Card of data acquisition, signal processing, positioning system, tempered system, pneumatic system, compute programs like it is the visual interface for the operation of the instrumented Charpy pendulum and the acquisition of impact signals. This system shows that given the characteristics of the nuclear industry with radioactive environments, the virtual instrumentation and the automation of processes can contribute to diminish the risks to the personnel occupationally exposed. (Author)

  17. Advanced Demonstration and Test Reactor Options Study

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy

  18. Advanced Demonstration and Test Reactor Options Study

    International Nuclear Information System (INIS)

    Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power's share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy's (DOE's) broader commitment to pursuing an 'all of the above' clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate 'advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy'. Advanced reactors are

  19. Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

    International Nuclear Information System (INIS)

    Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov; Joseph Palmer, A.; Teysseyre, Sebastien P.; Davis, Kurt L.; Rempe, Joy L.

    2015-01-01

    One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example. Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the

  20. On-line reactor building integrity testing at Gentilly-2 (summary of results 1987-1994)

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1994-01-01

    In 1987, Hydro-0uebec embarked on an ambitious development program to provide the Gentilly-2 Nuclear Power Station with an effective and practical Reactor Building Containment integrity Test (CIT). In October 1992, the inaugural low pressure (3 kPa(g) nominal) CIT at 100% F.P was performed. The test was conclusive and the CIT was declared In-Service for containment integrity verification on-line. Five subsequent CITs performed in 1993 and 1994 have demonstrated the expected leak rate results and good reliability. The outstanding feature of the CITs is the demonstrated accurary of better than 5% of the measured leak rate. The CIT was developed with the primary goal of demonstrating 'overall' containment availability. Specifically it was designed to detect a 25 mm. diameter leak or hole in the Reactor Building. However, the remarkable CIT accuracy allows reliable detection of a 2 mm. hole. The Gentilly-2 CIT is an innovative approach based on the Temperature Compensation Method (TCM) which uses a reference volume composed of an extensive tubular network of several different diameters. This eliminates the need to track numerous temperature points. A second independent tubular network includes numerous humidity sampling points, thereby enabling the mearurernent of minute pressure variations inside the Reactor Building, independant of the spatial and temporal humidity behaviour. This Gentilly-2 TOM System has been demonstrated to work at both high and low test pressures. The GentiIly-2 design allows the CIT to be performed at a nominal 3 kPa(g) test pressure during a 12-hour period (28 hours total with alignment time) with the reactor at full power. The traditional Reactor Building Pressure Test (RBPT) is typically performed at high pressure (124 kPa(g) in a 5-day critical path window (7 days total with alignment time) during an annual shutdown

  1. Refurbishing the BR2 materials testing reactor

    International Nuclear Information System (INIS)

    Baugnet, J.M.; Dekeyser, J.; Gubel, P.

    1995-01-01

    SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

  2. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Science.gov (United States)

    Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

    2008-08-01

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  3. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2008-08-31

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  4. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    OpenAIRE

    KO, DO-YOUNG; KIM, KYU-HYUNG

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

  5. The U.S. DOE new production reactor/heavy water reactor facility pollution prevention/waste minimization program

    International Nuclear Information System (INIS)

    Kaczmarsky, Myron M.; Tsang, Irving; Stepien, Walter P.

    1992-01-01

    A Pollution Prevention/Waste Minimization Program was established during the early design phase of the U.S. DOE's New Production Reactor/Heavy Water Reactor Facility (NPR/HWRF) to encompass design, construction, operation and decommissioning. The primary emphasis of the program was given to waste elimination, source reduction and/or recycling to minimize the quantity and toxicity of material before it enters the waste stream for treatment or disposal. The paper discusses the regulatory and programmatic background as it applies to the NPR/HWRF and the waste assessment program developed as a phased approach to pollution prevention/waste minimization for the NPR/HWRF. Implementation of the program will be based on various factors including life cycle cost analysis, which will include costs associated with personnel, record keeping, transportation, pollution control equipment, treatment, storage, disposal, liability, compliance and oversight. (author)

  6. SYNBURN: fast-reactor fuel-cycle program

    International Nuclear Information System (INIS)

    Pizzica, P.A.; Meneley, D.A.

    1976-01-01

    The SYNBURN computer program for fast reactors will calculate all the neutronics necessary to completely characterize the equilibrium cycle as well as the startup to equilibrium cycles. The program's run time is very short and this makes the program suitable for survey of parametric studies. It can search on the cycle time for a specified burnup, for the shim control necessary for criticality as well as feed enrichments and the enrichment ratio among core zones. SYNBURN synthesizes in a very simple fashion the one-dimensional fluxes in radial and axial geometry to achieve an approximate two-dimensional solution which agrees very well with the exact two-dimensional solution when measuring regional integrated quantities

  7. Computerised programming of the Dragon reactor fuel handling operations

    International Nuclear Information System (INIS)

    Butcher, P.

    1976-11-01

    Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

  8. An overview of the Nuclear Electric Propulsion Space Test Program (NEPSTP) satellite

    International Nuclear Information System (INIS)

    Voss, S.S.; Reynolds, E.L.

    1994-01-01

    Early in 1992 the idea of purchasing a Russian designed and fabricated space reactor power system and integrating it with a US designed satellite went from fiction to reality with the purchase of the first two Topaz II reactors by the Strategic Defense Initiative Organization (now the Ballistic Missile Defense Organization (BMDO)). The New Mexico Alliance was formed to establish a ground test facility in which to perform nonnuclear systems testing of the Topaz II, and to evaluate the Topaz II system for flight testing with respect to safety, performance, and operability. In conjunction, SDIO requested that the Applied Physics Laboratory in Laurel, MD propose a mission and design a satellite in which the Topaz II could be used as the power source. The outcome of these two activities was the design of the Nuclear Electric Propulsion Space Test Program (NEPSTP) satellite which combines a modified Russian Topaz II power system with a US designed satellite to achieve a specified mission. Due to funding reduction within the SDIO, the Topaz II flight program was postponed indefinitely at the end of Fiscal year 1993. The purpose of this paper is to present an overview of the NEPSTP mission and the satellite design at the time the flight program ended

  9. Application of the REMIX thermal mixing calculation program for the Loviisa reactor

    International Nuclear Information System (INIS)

    Kokkonen, I.; Tuomisto, H.

    1987-08-01

    The REMIX computer program has been validated to be used in the pressurized thermal shock study of the Loviisa reactor pressure vessel. The program has been verified against the data from the thermal and fluid mixing experiments. These experiments have been carried out in Imatran voima Oy to study thermal mixing of the high-pressure safety injection water in the Loviisa VVER-440 type pressurized water reactor. The verified REMIX-versions were applied to reactor calculations in the probabilistic pressurized thermal shock study of the Loviisa Plant

  10. Minutes of the 13th light water reactor pressure vessel surveillance dosimetry improvement program (LWR-PV-SDIP) meeting

    International Nuclear Information System (INIS)

    1984-04-01

    Information is presented concerning ASTM LWR standards and program documentation; trend curves, PSF, and other test reactor metallurgical programs; PSF dosimetry and metallurgical capsule neutron and gamma environment characterization and metallurgical studies; PVS characterization program; other neutron fields; surveillance dosimetry measurement facility (SDMF) and perturbation studies; transport theory calculations; gamma field benchmarks and photo-reaction studies; and fission and non-fission sensor inventories and quality assurance

  11. A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodiac, F.; Hudelot, JP.; Lecerf, J.; Garnier, Y.; Ritter, G. [CEA, DEN, CAD/DER/SRES/LPRE, Cadarache, F-13108 Saint-Paul-lez-Durance, (France); Gueton, O.; Colombier, AC. [CEA, DEN, CAD/DER/SPRC/LPN, Cadarache, F-13108 Saint-Paul-lez-Durance, (France); Domergue, C. [CEA, DEN, CAD/DER/SPEx/LDCI, Cadarache, F-13108 Saint-Paul-lez-Durance, (France)

    2015-07-01

    CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimental program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)

  12. Light Water Reactor Sustainability Program: Evaluation of Localized Cable Test Methods for Nuclear Power Plant Cable Aging Management Programs

    Energy Technology Data Exchange (ETDEWEB)

    Glass, Samuel W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fifield, Leonard S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hartman, Trenton S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-30

    This Pacific Northwest National Laboratory (PNNL) milestone report describes progress to date on the investigation of nondestructive test (NDE) methods focusing particularly on local measurements that provide key indicators of cable aging and damage. The work includes a review of relevant literature as well as hands-on experimental verification of inspection capabilities. As NPPs consider applying for second, or subsequent, license renewal (SLR) to extend their operating period from 60 years to 80 years, it important to understand how the materials installed in plant systems and components will age during that time and develop aging management programs (AMPs) to assure continued safe operation under normal and design basis events (DBE). Normal component and system tests typically confirm the cables can perform their normal operational function. The focus of the cable test program is directed toward the more demanding challenge of assuring the cable function under accident or DBE. Most utilities already have a program associated with their first life extension from 40 to 60 years. Regrettably, there is neither a clear guideline nor a single NDE that can assure cable function and integrity for all cables. Thankfully, however, practical implementation of a broad range of tests allows utilities to develop a practical program that assures cable function to a high degree. The industry has adopted 50% elongation at break (EAB) relative to the un-aged cable condition as the acceptability standard. All tests are benchmarked against the cable EAB test. EAB is a destructive test so the test programs must apply an array of other NDE tests to assure or infer the overall set of cable’s system integrity. These cable NDE programs vary in rigor and methodology. As the industry gains experience with the efficacy of these programs, it is expected that implementation practice will converge to a more common approach. This report addresses the range of local NDE cable tests that are

  13. In-core materials testing under LWR conditions in the Halden reactor

    International Nuclear Information System (INIS)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

    2002-01-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  14. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  15. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  16. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  17. Situation of test and research reactors' spent fuels

    International Nuclear Information System (INIS)

    Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

    1996-01-01

    The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

  18. MSR - SPHINX concept program Eros (Experimental zero power Salt reactor SR-0) - The proposed experimental program as a basis for validation of reactor physics methods

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Juricek, V.; Kyncl, J.; Mikisek, M.; Rypar, V. [Nuclear Research Institute Rez plc, Rez (Czech Republic)

    2007-07-01

    The Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept solves this principal problem of spent fuel treatment by means of so-called nuclear incineration. It means the burning of fissionable part of its inventory and transmutation of other problematic radionuclides by use of nuclear reactions with neutrons in a MSR-SPHINX system. This reactor system is an actinide burner (most in resonance neutron spectrum) and a radionuclide transmuter in a well-thermalized neutron spectrum. In the frame of the physical part, there are computational analyses and experimental activities. The experimental program has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors. The proposed next stage of the program will focus on a large-scale experimental verification of design inputs by use of MSR-type inserting zones into the existing light water moderated experimental reactor LR-0, which may allow us to modify it into the experimental zero power salt reactor SR-0. There will be a detail description of the proposed program given in the paper together with the so far performed experiments and their first results. These realized experiments help us also to verify computational codes used, and to recognize some anomalies related to molten fluorides utilization. (authors)

  19. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    Energy Technology Data Exchange (ETDEWEB)

    A. B. Culp

    2007-01-26

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

  20. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

  1. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

  2. Quality assurance program for surveillance of fast reactor mixed oxide fuel analytical chemistry

    International Nuclear Information System (INIS)

    Rein, J.E.; Zeigler, R.K.; Waterbury, G.R.; McClung, W.E.; Praetorius, P.R.; Delvin, W.L.

    1976-01-01

    An effective quality assurance program for the chemical analysis of nuclear fuel is essential to assure that the fuel will meet the strict chemical specifications required for optimum reactor performance. Such a program has been in operation since 1972 for the fuels manufactured for the Fast Flux Test Facility. This program, through the use of common quality control and calibration standards, has consistently provided high levels of agreement among laboratories in all areas of analysis. The paper presented gives a summary of the chemical specifications for the fuel and source material, an outline of the requirements for laboratory qualifications and the preparation of calibration and quality control materials, general administration details of the plan, and examples where the program has been useful in solving laboratory problems

  3. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  4. Advanced Test Reactor probabilistic risk assessment

    International Nuclear Information System (INIS)

    Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

    1993-01-01

    This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

  5. Thermal-hydraulic tests for reactor safety system

    International Nuclear Information System (INIS)

    Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

    2002-05-01

    Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

  6. State of the reactor vessel surveillance programs in Korea and foreign countries

    International Nuclear Information System (INIS)

    Kim, Jeong Kyu; Hwang, Jong Keun; Park, Keon Woo; Kim, Bum Sik; Jeong, Kyung Hoon

    1996-06-01

    ASTM standards are dominating all over the world in the field of the reactor vessel surveillance program. They are mainly used directly or that the national standards in use correspond quite well with ASTM. According to, however, increasing concerns about the protection of environment and safety of nuclear plant, various approaches to establish and reinforce the national standards are made actively in Europe. In addition, some methods to share the nuclear data by integrating the existing test, analysis procedures and units system are considered. For nuclear plants in Korea, MOST Notice No. 92-20 should be applied for all PWRs after UCN units 3 and 4 since it was promulgated at Dec. 1992. The notice almost reflects the contents of ASTM E 185. But, the notice has much to be desired to provide the technical back-ground for reactor vessel surveillance program because it is not a standard such as ASTM or ASME code but regulation such as CFR or RG. Several Korean Standards are also used in limited area of the surveillance program. Therefore, practical requirements and rules for surveillance program are in accordance with the ASTM and CFR. In this report, the state of application of the standards to the surveillance program in Korea and Europe are reviewed and their national standards re compared with US standards or regulations. Current level and the future prospect of surveillance technology for PWR vessel are discussed at this point of view. 15 tabs., 12 figs., 38 refs. (Author)

  7. Potential of duplex fuel in prebreeder, breeder, and power reactor designs: tests and analyses (AWBA Development Program)

    International Nuclear Information System (INIS)

    Chao, T.L.; Brennan, J.J.; Duncombe, E.; Schneider, M.J.; Johnson, R.G.R.

    1982-09-01

    Dual region fuel pellets, called duplex pellets, are comprised of an outer annular region of relatively high uranium fuel enrichment and a center pellet of fertile material with no enrichment. UO 2 and ThO 2 are the fissile and fertile materials of interest. Both prebreeders and breeders are discussed as are the performance advantages of duplex pellets over solid pellets in these two pressurized water reactor types. Advantages of duplex pellets for commercial reactor fuel rods are also discussed. Both irradiation test data and analytical results are used in comparisons. Manufacturing of duplex fuel is discussed

  8. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  9. Evaluation of the qualification of SPERT [Special Power Excursion Reactor Test] fuel for use in non-power reactors

    International Nuclear Information System (INIS)

    1987-08-01

    This report summarizes the US Nuclear Regulatory Commission staff's evaluation of the qualification of the stainless-steel-clad uranium/oxide (UO 2 ) fuel pins for use in non-power reactors. The fuel pins were originally procured in the 1960's as part of the Special Power Excursion Reactor Test (SPERT) program. Argonne National Laboratory (ANL) examined 600 SPERT fuel pins to verify that the pins were produced according to specification and to assess their present condition. The pins were visually inspected under 6X magnification and by X-radiographic, destructive, and metallographic examinations. Spectrographic and chemical analyses were performed on the UO 2 fuel. The results of the qualification examinations indicated that the SPERT fuel pins meet the requirements of Phillips Specification No. F-1-SPT and have suffered no physical damage since fabrication. Therefore, the qualification results give reasonable assurance that the SPERT fuel rods are suitable for use in non-power reactors provided that the effects of thin-wall defects in the region of the upper end cap and low-density fuel pellets are evaluated for the intended operating conditions. 1 ref., 4 figs., 11 tabs

  10. A two-step method for developing a control rod program for boiling water reactors

    International Nuclear Information System (INIS)

    Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

    1992-01-01

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift

  11. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  12. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    International Nuclear Information System (INIS)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately

  13. Reactor materials program process water component failure probability

    International Nuclear Information System (INIS)

    Daugherty, W. L.

    1988-01-01

    The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system

  14. Treatment of spent fuels from research reactors and reactor development programs in Germany

    International Nuclear Information System (INIS)

    Closs, K.D.

    1999-01-01

    Quite a great number of different types of spent fuel from research reactors and development programs exists in Germany. The general policy is to send back to the USA as long as possible fuel from MTRs and TRIGAs of USA origin. An option is reprocessing in Great Britain or France. This option is pursued as long as reprocessing and reuse of the recovered material is economically justifiable. For those fuels which cannot be returned to the USA or which will not be reprocessed, a domestic back-up solution of spent fuel management has been developed in Germany, compatible with the management of spent fuel from power reactors. It consists in dry storage in special casks and, later on, direct disposal. Preliminary results from experimental R and D investigations with research reactor fuel and experience from LWR fuel lead to the conclusion that the direct disposal option even for research reactor fuel or exotic fuel does not impose major technical difficulties for the German waste management and disposal concept. (author)

  15. Progress of the United States foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Huizenga, D.G.; Clapper, M.; Thrower, A.W.

    2002-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program has completed 23 shipments. Almost 5000 spent fuel assemblies from eligible research reactors throughout the world have been accepted into the United States under this program. Over the past year, another cross-country shipment of fuel was accomplished, as well as two additional shipments in the fourth quarter of calendar year 2001. These shipments attracted considerable safeguards oversight since they occurred post September 11. Recent guidance from the Nuclear Regulatory Commission (NRC) pertaining to security and safeguards issues deals directly with the transport of nuclear material. Since the Acceptance Program has consistently applied above regulatory safety enhancements in transport of spent nuclear fuel, this guidance did not adversely effect the Program. As the Program draws closer to its termination date, an increased number of requests for program extension are received. Currently, there are no plans to extend the policy beyond its current expiration date; therefore, eligible reactor operators interested in participating in this program are strongly encouraged to evaluate their inventory and plan for future shipments as soon as possible. (author)

  16. Progress and status of the integral fast reactor (IFR) development program

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    This paper discusses the Integral Fast Reactor (IFR) development program, in which the entire reactor system - reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are also presented

  17. The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

    International Nuclear Information System (INIS)

    Bruschi, H.J.

    1996-01-01

    At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse's advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

  18. The AP600 advanced simplified nuclear power plant. Results of the test program and progress made toward final design approval

    Energy Technology Data Exchange (ETDEWEB)

    Bruschi, H.J. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1996-10-01

    At the 1994 Pacific Basin Conference, Mr. Bruschi presented a paper describing the AP600, Westinghouse`s advanced light water reactor design with passive safety features. Since then, a rigorous test program was completed and AP600 became the most thoroughly tested advanced reactor system design in history. Westinghouse is now well on its way toward receiving Final Design Approval from the U.S. Nuclear Regulatory Commission for AP600. In this paper, the results of the test program will be discussed and an update on prospects for building the plant will be covered. (author)

  19. Investigation of the loss of forced cooling test by using the high temperature engineering test reactor (HTTR) (Contract research)

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Inaba, Yoshitomo; Goto, Minoru; Tochio, Daisuke

    2007-09-01

    The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur. (author)

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

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2008-01-01

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

  1. Participation in the United States Department of Energy Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993.

  2. Participation in the United States Department of Energy Reactor Sharing Program

    International Nuclear Information System (INIS)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1992-05-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics (to become the Department of Mechanical, Aerospace and Nuclear Engineering on July 1, 1992). As such, it is effectively used to support educational programs in engineering and science at the University of Virginia as well as those at other area colleges and universities. The expansion of support to educational programs in the mid-east region is a major objective. To assist in meeting this objective, the University of Virginia has been supported under the US Department of Energy (DOE) Reactor Sharing Program since 1978. Due to the success of the program, this proposal requests continued DOE support through August 1993

  3. Refurbish research and test reactors corresponding to global age of nuclear energy

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Oyama, Yukio; Okamoto, Koji; Yamana, Hajime; Yamaguchi, Akira

    2011-01-01

    This special article featured arguments for refurbishment of research and test reactors corresponding to global age of nuclear energy, based on the report: 'Investigation of research facilities necessary for future joint usage' issued by the special committee of Atomic Energy Society of Japan (AESJ) in September 2010. It consisted of six papers titled as 'Introduction-establishment of AESJ special committee for investigation', 'State of research and test reactors in Japan', 'State of overseas research and test reactors', 'Needs analysis for research and test reactors', 'Proposal of AESJ special committee' and 'Summary and future issues'. In order to develop human resources and promote research and development needed in global age of nuclear energy, research and test reactors would be refurbished as an Asian regional center of excellence. (T. Tanaka)

  4. Reactor cover gas monitoring at the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, R A; Holt, F E; Meadows, G E; Schenter, R E [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    The Fast Flux Test Facility (FFTF) is a 400 megawatt (thermal) sodium cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the U. S. Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100 day operating cycle began in April 1982 and the eighth operating cycle was completed In July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification. A liquid argon Dewar system provides the large volume of inert gas required for operation of the FFTF. The gas is used as received and is not recycled. Low concentrations of krypton and xenon in the argon supply are essential to preclude interference with the gas tag system. Gas chromatography has been valuable for detection of inadvertent air in leakage during refueling operations. A temporary system is installed over the reactor during outages to prevent oxide formation in the sodium vapor traps upstream from the on line gas chromatograph. On line gas monitoring by gamma spectrometry and grab sampling with GTSTs has been successful for the identification of numerous radioactive gas releases from creep capsule experiments as well as 9 fuel pin ruptures. A redundant fission gas monitoring system has been installed to insure constant surveillance of the reactor cover gas.

  5. Enhanced in-pile instrumentation at the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

  6. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    Science.gov (United States)

    Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

    2012-08-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  7. Decentralization of operating reactor licensing reviews: NRR Pilot Program

    International Nuclear Information System (INIS)

    Hannon, J.N.

    1984-07-01

    This report, which has incorporated comments received from the Commission and ACRS, describes the program for decentralization of selected operating reactor licensing technical review activities. The 2-year pilot program will be reviewed to verify that safety is enhanced as anticipated by the incorporation of prescribed management techniques and application of resources. If the program fails to operate as designed, it will be terminated

  8. International topical meeting on research reactor fuel management (RRFM) - United States foreign research reactor (FRR) spent nuclear fuel (SNF) acceptance program: 2010 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.; Niehus, M.T.; Landers, C.

    2010-01-01

    The Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by the United States Department of Energy (DOE), in consultation with the Department of State (DOS) in May 1996, scheduled to expire May 12, 2016, to return research reactor fuel until May 12, 2019 to the U.S. is in its fourteenth year. This paper provides a brief update on the program, part of the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of U.S.-origin nuclear materials, which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible research reactors who have not yet participated in the program. (author)

  9. Analyzing and comparing the dynamic response of test reactor main workshop

    International Nuclear Information System (INIS)

    Wang Jiachun; Fu Jiyang; Cai Laizhong

    2001-01-01

    Analyzing soil-structure interaction is an important section in anti-seismic design and analysis of nuclear engineering. The factors that influence on the response of nuclear structures include the properties of earthquake, soil and structures. So the soil-structure interaction in the non-rock foundation is different from that in the surface free field. And the interaction must be considered under the anti-seismic design standard of test reactors. The FLUSH program and SASSI2000 are applied to dynamic analysis. Moreover, comparing the obtained data and diagrams draws some conclusions

  10. Cavity temperature and flow characteristics in a gas-core test reactor

    Science.gov (United States)

    Putre, H. A.

    1973-01-01

    A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

  11. Instruments for non-destructive evaluation of advanced test reactor inpile tubes

    International Nuclear Information System (INIS)

    Livingston, R.A.; Beller, L.S.; Edgett, S.M.

    1986-01-01

    The Advanced Test Reactor is a 250 MW LWR used primarily for irradiation testing of materials contained in inpile tubes that pass through the reactor core. These tubes provided the high pressure and temperature water environment required for the test specimens. The reactor cooling water surrounding the inpile tubes is at much lower pressure and temperature. The structural integrity of the inpile tubes is monitored by routine surveillance to ensure against unplanned reactor shutdowns to replace defective inpile tubes. The improved instruments developed for inpile tube surveillance include a bore profilometer, ultrasonic flaw detetion system and bore diameter gauges. The design and function of these improved instruments is presented

  12. Overview and status of the SIMMER testing program

    International Nuclear Information System (INIS)

    Scott, J.H.

    1979-01-01

    Los Alamos Scientific Laboratory has undertaken an extensive experiment analysis program to test the results of SIMMER Liquid Metal Fast Breeder Reactor (LMFBR) accident calculations. Initially, we will test the postdisassembly work-energy partition problem. The SIMMER-calculated order-of-magnitude reduction of available kinetic energy following a severe hypothetical core-disruptive accident (HCDA) can be attributed to (1) purely fluid-dynamic effects; and (2) rate-controlled effects, such as phase transitions and heat transfer. We have chosen to test separately each class of mitigator. In this paper we review the experiments initially chosen for testing of each class of mitigator and report on the status of the analyses. We enumerate several problems in SIMMER that experiment analysis has disclosed. Finally, needs for future experiments are discussed

  13. Assessment of specialized educational programs for licensed nuclear reactor operators

    International Nuclear Information System (INIS)

    Melber, B.D.; Saari, L.M.; White, A.S.; Geisendorfer, C.L.; Huenefeld, J.C.

    1986-02-01

    This report assesses the job-relatedness of specialized educational programs for licensed nuclear reactor operators. The approach used involved systematically comparing the curriculum of specialized educational programs for college credit, to academic knowledge identified as necessary for carrying out the jobs of licenses reactor operators. A sample of eight programs, including A.S. degree, B.S. degree, and coursework programs were studied. Subject matter experts in the field of nuclear operations curriculum and training determined the extent to which individual program curricula covered the identified job-related academic knowledge. The major conclusions of the report are: There is a great deal of variation among individual programs, ranging from coverage of 15% to 65% of the job-related academic knowledge. Four schools cover at least half, and four schools cover less than one-third of this knowledge content; There is no systematic difference in the job-relatedness of the different types of specialized educational programs, A.S. degree, B.S. degree, and coursework; and Traditional B.S. degree programs in nuclear engineering cover as much job-related knowledge (about one-half of this knowledge content) as most of the specialized educational programs

  14. Reactor recirculation pump test loop

    International Nuclear Information System (INIS)

    Taka, Shusei; Kato, Hiroyuki

    1979-01-01

    A test loop for a reactor primary loop recirculation pumps (PLR pumps) has been constructed at Ebara's Haneda Plant in preparation for production of PLR pumps under license from Byron Jackson Pump Division of Borg-Warner Corporation. This loop can simulate operating conditions for test PLR pumps with 130 per cent of the capacity of pumps for a 1100 MWe BWR plant. A main loop, primary cooling system, water demineralizer, secondary cooling system, instrumentation and control equipment and an electric power supply system make up the test loop. This article describes the test loop itself and test results of two PLR pumps for Fukushima No. 2 N.P.S. Unit 1 and one main circulation pump for HAZ Demonstration Test Facility. (author)

  15. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

    Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

  16. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

    International Nuclear Information System (INIS)

    2005-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)

  17. The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

    Science.gov (United States)

    Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

    2018-01-01

    The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

  18. Final report. U.S. Department of Energy University Reactor Sharing Program

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, John A

    2003-01-21

    Activities supported at the MIT Nuclear Reactor Laboratory under the U.S. DOE University Reactor Sharing Program are reported for Grant DE FG02-95NE38121 (September 16, 1995 through May 31, 2002). These activities fell under four subcategories: support for research at thesis and post-doctoral levels, support for college-level laboratory exercises, support for reactor tours/lectures on nuclear energy, and support for science fair participants.

  19. Reactor console replacement at Washington State University

    International Nuclear Information System (INIS)

    Lovas, Thomas A.

    1978-01-01

    A replacement reactor console was installed in 1977 at the W.S.U. 1 MW TRIGA-fueled reactor as the final step in an instrumentation upgrade program. The program was begun circa 1972 with the design, construction and installation of various systems and equipment. Major instruments were installed in the existing console and tested in the course of reactor operation. The culmination of the program was the installation of a cubicle designed and constructed to house the updated instrumentation. (author)

  20. Role of fission-reactor-testing capabilities in the development of fusion technology

    International Nuclear Information System (INIS)

    Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

    1981-01-01

    Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

  1. The US Liquid-Metal Reactor Program - overview and status

    International Nuclear Information System (INIS)

    Quinn, J.E.; Gyorey, G.L.; Salerno, L.N.

    1992-01-01

    The US Advanced Liquid-Metal Reactor (ALMR) Program has three major elements being developed in an integrated fashion to produce a system meeting the US long-term nuclear energy needs. Reactor design, one of those elements, is the focus of this paper. The other two elements, the integral fast reactor metal-fuel cycle and the light water reactor (LWR) spent-fuel actinide recycle, will be addressed in companion papers. The ALMR is adaptable to multiple missions with few modifications such as the core arrangements. The missions identified to date are (a) the extension of the existing uranium resources through breeding and highly efficient uranium utilization, (b) the recycle and utilization of the long-life actinides in LWR spent fuel as fissile material for the ALMR, and (c) the conversion of excess weapons fissil material into electricity. In addition to these missions, the reactor design is adaptable to either the metal-fuel cycle or the oxide fuel cycle

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. Progress and status of the Integral Fast Reactor (IFR) development program

    International Nuclear Information System (INIS)

    Chang, Yoon I.

    1992-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this

  4. Progress and status of the Integral Fast Reactor (IFR) development program

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1992-04-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this.

  5. Progress and status of the Integral Fast Reactor (IFR) development program

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Yoon I.

    1992-01-01

    In the Integral Fast Reactor (IFR) development program, the entire reactor system -- reactor, fuel cycle, and waste process is being developed and optimized at the same time as a single integral entity. The ALMR reactor plant design is being developed by an industrial team headed by General Electric and is presented in a companion paper. Detailed discussions on the present status of the IFR technology development activities in the areas of fuels, pyroprocessing, safety, core design, and fuel cycle demonstration are presented in the other two companion papers that follows this.

  6. The High Level Vibration Test program

    International Nuclear Information System (INIS)

    Hofmayer, C.H.; Curreri, J.R.; Park, Y.J.; Kato, W.Y.; Kawakami, S.

    1990-01-01

    As part of cooperative agreements between the United States and Japan, tests have been performed on the seismic vibration table at the Tadotsu Engineering Laboratory of Nuclear Power Engineering Test Center (NUPEC) in Japan. The objective of the test program was to use the NUPEC vibration table to drive large diameter nuclear power piping to substantial plastic strain with an earthquake excitation and to compare the results with state-of-the-art analysis of the problem. The test model was designed by modifying the 1/2.5 scale model of the pressurized water reactor primary coolant loop. Elastic and inelastic seismic response behavior of the test model was measured in a number of test runs with an increasing excitation input level up to the limit of the vibration table. In the maximum input condition, large dynamic plastic strains were obtained in the piping. Crack initiation was detected following the second maximum excitation run. The test model was subjected to a maximum acceleration well beyond what nuclear power plants are designed to withstand. This paper describes the overall plan, input motion development, test procedure, test results and comparisons with pre-test analysis

  7. Nuclear reactor development in China for non-electrical applications

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Dong Duo; Xu Yuanhui

    1998-01-01

    In parallel to its vigorous program of nuclear power generation, China has attached great importance to the development of nuclear reactors for non-electrical applications. The Institute of Nuclear Energy Technology (INET) in Beijing has been developing technologies of the water-cooled heating reactor and the modular high temperature gas-cooled reactor. In 1989, a 5 MW water cooled test reactor was erected. Currently, an industrial demonstration nuclear heating plant is being projected. Feasibility studies are being made of sea-water desalination using the INET developed nuclear heating reactor as heat source. Also, a 10 MW high temperature gas-cooled test reactor is being constructed at INET in the framework of China's national high-tech program. The paper gives an overview of China's energy market situation. With respect to China's technology development of high temperature gas-cooled reactors and water cooled heating reactors, the paper describes some general requirements on the technical development, reviews the national programs and activities, describes briefly the design and safety features of the reactor concepts, discusses aspects of application potentials. (author)

  8. Participation in the U.S. Department of Energy Reactor Sharing Program. Progress report

    International Nuclear Information System (INIS)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here

  9. Needs for development in nondestructive testing for advanced reactor systems

    International Nuclear Information System (INIS)

    McClung, R.W.

    1978-01-01

    The needs for development of nondestructive testing (NDT) techniques and equipment were surveyed and analyzed relative to problem areas for the Liquid-Metal Fast Breeder Reactor, the Molten-Salt Breeder Reactor, and the Advanced Gas-Cooled Reactor. The paper first discusses the developmental needs that are broad-based requirements in nondestrutive testing, and the respective methods applicable, in general, to all components and reactor systems. Next, the requirements of generic materials and components that are common to all advanced reactor systems are examined. Generally, nondestructive techniques should be improved to provide better reliability and quantitativeness, improved flaw characterization, and more efficient data processing. Specific recommendations relative to such methods as ultrasonics, eddy currents, acoustic emission, radiography, etc., are made. NDT needs common to all reactors include those related to materials properties and degradation, welds, fuels, piping, steam generators, etc. The scope of applicability ranges from initial design and material development stages through process control and manufacturing inspection to in-service examination

  10. The United States foreign research reactor spent nuclear fuel acceptance program: Proposal to modify the program

    International Nuclear Information System (INIS)

    Messick, C.E.

    2005-01-01

    The United States Department of Energy (DOE), in consultation with the Department of State (DOS), adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. The policy was slated to expire in May 2009. However, in October 2003, a petition requesting a program extension was delivered to the United States Secretary of Energy from a group of research reactor operators from foreign countries. In April 2004, the Secretary directed DOE undertake an analysis, as required by the National Environmental Policy Act (NEPA), to consider potential extension of the Program. On December 1, 2004, a Federal Register Notice was issued approving the program extension. This paper discusses the findings from the NEPA analysis and the potential changes in the program that may result from implementation of the proposed changes. (author)

  11. Service to the Electric Utility Industry by the Ford Nuclear Reactor, University of Michigan

    International Nuclear Information System (INIS)

    Burn, R.R.; Simpson, P.A.; Cook, G.M.

    1993-01-01

    Since 1977, the staff of the University of Michigan's Ford Nuclear Reactor has been providing irradiation, testing, analytical, and training services to electric utilities and to suppliers of the nuclear electric utility industry. This paper discusses the reactor's irradiation facilities; reactor programs and utilization; materials testing programs; neutron activation analysis activities; and training programs conducted

  12. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

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

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

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

  14. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    International Nuclear Information System (INIS)

    McCarthy, K.A.; Williams, D.L.; Reister, R.

    2012-01-01

    The US Department of Energy Light Water Reactor Sustainability (LWRS) Program is focused on enabling the long-term operation of US commercial power plants. Decisions on life extension will be made by commercial power plant owners - the information provided by the research and development activities in the LWRS Program will reduce the uncertainty (and therefore the risk) associated with making those decisions. The LWRS Program encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper provides an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables. (author)

  15. A structured approach to evaluating aging of the advanced test reactor

    International Nuclear Information System (INIS)

    Dwight, J.E.

    1990-01-01

    An aging evaluation program has been developed for the United States Department of Energy's Advanced Test Reactor to support the current goal of operation through the year 2014 and beyond. The Aging Evaluation and Life Extension Program (AELEX) employs a three-phased approach. In Phases 1 and 2, now complete, components were identified, categorized and prioritized. Critical components were selected and aging mechanisms for the critical components identified. An initial evaluation of the critical components was performed and extended life operation for the plant appears to be both technically and economically feasible. Detailed evaluations of the critical components are now in progress in the early stages of Phase 3. Some results are available. Evaluations of many non-critical components and refinements to the program based on probabilistic risk assessment results will follow in later stages of Phase 3. 6 refs., 2 figs., 5 tabs

  16. PWR Blowdown Heat Transfer Separate-Effects Program. Thermal-Hydraulic Test Facility experimental data report for test 166S

    International Nuclear Information System (INIS)

    Clemons, V.D.; White, M.D.; Hedrick, R.A.

    1978-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 166S, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 166S was conducted to obtain thermal-hydraulic and CHF information in THTF bundle 1 with an intact hot leg. The primary purpose of this report is to make the reduced instrument responses during tests 166S available. These are presented in graphical form in engineering units and have been analyzed only to the extent necessary to ensure reasonableness and consistency

  17. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

    International Nuclear Information System (INIS)

    2016-02-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

  18. Safety Technology Research Program in the field of pressurized water reactors. 1. Technical report on advancement project RS 36/2. Emergency cooling program service life experiments: reflooding experiments involving the primary loop systems

    International Nuclear Information System (INIS)

    Schweickert, H.; Kremin, H.; Mandl, R.; Riedle, V.; Ruthrof, K.; Sarkar, J.; Schmidt, H.

    The reflooding of the hot reactor core is to be examined for a pressurized water reactor (PWR), using a model of the entire primary loop system. The scale of the model is to be 1:340 in cross-section, with the heights represented full-scale. In addition to the goals of the project, a description of the test facility, including data collection and control equipment is presented. The instrumentation, the planned test program and the test procedure are briefly set forth

  19. Tensile and impact testing of an HFBR [High Flux Beam Reactor] control rod follower

    International Nuclear Information System (INIS)

    Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

    1989-08-01

    The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (K I ) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K max values of 24.8 ± 1.0 Ksi√in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs

  20. Analysis and testing of W-DHR system for decay heat removal in the lead-cooled ELSY reactor

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Meloni, Paride; Polidori, Massimiliano; Gaggini, Piero; Labanti, Valerio; Tarantino, Mariano; Cinotti, Luciano; Presciuttini, Leonardo

    2009-01-01

    An innovative LFR system that complies with GEN IV goals is under design in the frame of ELSY European project. ELSY is a lead-cooled pool-type reactor of about 1500 MW thermal power which normally relies on the secondary system for decay heat removal. Since the secondary system is not safety-grade and must be fully depressurized in case of detection of a steam generator tube rupture, an independent and much reliable decay heat removal (DHR) system is foreseen on the primary side. Owing to the limited capability of the Reactor Vessel Air Cooling System (RVACS) in this large power reactor, additional safety-grade loops equipped with coolers immersed in the primary coolant are necessary for an efficient removal of decay heat. Some of these loops (W-DHR) are of innovative design and may operate with water at atmospheric pressure. In the frame of the ICE program to be performed on the integral facility CIRCE at ENEA/Brasimone research centre within the EUROTRANS European project, integral circulation experiments with core heat transport and heat removal by steam generator will be conducted in a reactor pool-type configuration. Taking advantage from this experimental program, a mock-up of W-DHR heat exchanger will be tested in order to investigate its functional behavior for decay heat removal. Some pre-test calculations of W-DHR heat exchanger operation in CIRCE have been performed with the RELAP5 thermal-hydraulic code in order to support the heat exchanger design and test conduct. In this paper the experimental activity to be conducted in CIRCE and main results from W-DHR pre-test calculations are presented, along with a preliminary investigation of the W-DHR system efficiency in ELSY configuration. (author)

  1. CRBRP flow induced vibration program

    Energy Technology Data Exchange (ETDEWEB)

    Novendstern, E H [Westinghouse Advanced Reactor Division, Madison, PA (United States); Grochowski, F A; Yang, T M [General Electric Co., Fast Breeder Reactor Department, Sunnyvale, CA (United States); Ryan, J A; Mulcahy, T M

    1977-12-01

    The program to assure the structural adequacy of Clinch River Breeder Reactor (CRBRP) components during its planned 30 years of operation is described. The program includes (1) an assessment of reactor components relative to their susceptibility to FIV, (2) designing to minimize component excitation due to Fluid induced vibrations (FIV), (3) scale model tests to measure structural response during simulated operating conditions and (4) preoperational tests. An overview of the CRBRP test program is described. Additionally, details of scale model testing of reactor internals and the steam generator is described in more detail. (author)

  2. PWR blowdown heat transfer separate-effects program: Thermal-Hydraulic Test Facility experimental data report for test 100

    International Nuclear Information System (INIS)

    White, M.D.; Hedrick, R.A.

    1977-01-01

    Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 100, which is part of the ORNL Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in a PWR system. Test 100 was conducted to investigate the response of heater rod bundle 1 and instrumented spool pieces with flow homogenizing screens to a double-ended rupture with equal break areas at the test section inlet and outlet. The primary purpose of this report is to make the reduced instrument responses during test 100 available. The responses are presented in graphical form in engineering units and have been analyzed only to the extent necessary to assure reasonableness and consistency

  3. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    International Nuclear Information System (INIS)

    Salam, M. A.

    2013-01-01

    by the installation of a high performance neutron powder diffractometer at the radial beam port-2 of the reactor. Although preventive and predictive maintenance, periodic testing, in-service inspection and other mitigation measures were practiced in few systems of the BAEC RR facility, these were not enough to mitigate the ageing related degradation of N-16 decay tank, shell and tube heat exchanger and radial beam port-1. It was understood that if effective ageing management program were started effectively from the beginning of the reactor, refurbishment activities could be minimized. As well as life of the reactor could be enhanced without any hurdle tasks. Basis on these understanding ROMU submitted an IAEA TC Project (BGD1012) titled 'Implementing an Ageing Management Program for RR'. Under this project an effective ageing management program has commenced on January, 2012 for the BAEC RR to enhance life and ensure safe operation of the reactor. This paper will describe chronologically refurbishment activities and experience for different affected systems, required resources and management and lessons learned from these tasks. It will also focus the modernization activities of the reactor facility. Moreover, it will describe ageing management activities that are planned for the BAEC TRIGA MK-II research reactor

  4. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Salam, M. A. [Atomic Energy Research Establishment, Dhaka (Bangladesh)

    2013-07-01

    by the installation of a high performance neutron powder diffractometer at the radial beam port-2 of the reactor. Although preventive and predictive maintenance, periodic testing, in-service inspection and other mitigation measures were practiced in few systems of the BAEC RR facility, these were not enough to mitigate the ageing related degradation of N-16 decay tank, shell and tube heat exchanger and radial beam port-1. It was understood that if effective ageing management program were started effectively from the beginning of the reactor, refurbishment activities could be minimized. As well as life of the reactor could be enhanced without any hurdle tasks. Basis on these understanding ROMU submitted an IAEA TC Project (BGD1012) titled 'Implementing an Ageing Management Program for RR'. Under this project an effective ageing management program has commenced on January, 2012 for the BAEC RR to enhance life and ensure safe operation of the reactor. This paper will describe chronologically refurbishment activities and experience for different affected systems, required resources and management and lessons learned from these tasks. It will also focus the modernization activities of the reactor facility. Moreover, it will describe ageing management activities that are planned for the BAEC TRIGA MK-II research reactor.

  5. Reactor Physics Experiments by Korean Under-Graduate Students in Kyoto University Critical Assembly Program (KUGSiKUCA Program)

    International Nuclear Information System (INIS)

    Pyeon, Cheol Ho; Misawa, Tsuyoshi; Unesaki, Hironobu; Ichihara, Chihiro; Shiroya, Seiji; Whang, Joo Ho; Kim, Myung Hyun

    2006-01-01

    The Reactor Laboratory Course for Korean Under-Graduate Students in Kyoto University Critical Assembly (KUGSiKUCA) program has been launched from 2003, as one of international collaboration programs of Kyoto University Research Reactor Institute (KURRI). This program was suggested by Department of Nuclear Engineering, College of Advanced Technology, Kyunghee University (KHU), and was adopted by Ministry of Science and Technology of Korean Government as one of among Nuclear Human Resources Education and Training Programs. On the basis of her suggestion for KURRI, memorandum for academic corporation and exchange between KHU and KURRI was concluded on July 2003. The program has been based on the background that it is extremely difficult for any single university in Korea to have her own research or training reactor. Up to this 2006, total number of 61 Korean under-graduate school students, who have majored in nuclear engineering of Kyunghee University, Hanyang University, Seoul National University, Korea Advanced Institute of Science and Technology, Chosun University and Cheju National University in all over the Korea, has taken part in this program. In all the period, two professors and one teaching assistant on the Korean side led the students and helped their successful experiments, reports and discussions. Due to their effort, the program has succeeded in giving an effective and unique course, taking advantage of their collaboration

  6. International benchmark on the natural convection test in Phenix reactor

    International Nuclear Information System (INIS)

    Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

    2013-01-01

    Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

  7. Thermal Hydraulic Fortran Program for Steady State Calculations of Plate Type Fuel Research Reactors

    International Nuclear Information System (INIS)

    Khedr, H.

    2008-01-01

    The safety assessment of Research and Power Reactors is a continuous process over their life and that requires verified and validated codes. Power Reactor codes all over the world are well established and qualified against a real measuring data and qualified experimental facilities. These codes are usually sophisticated, require special skills and consume much more running time. On the other hand, most of the Research Reactor codes still requiring more data for validation and qualification. Therefore it is benefit for a regulatory body and the companies working in the area of Research Reactor assessment and design to have their own program that give them a quick judgment. The present paper introduces a simple one dimensional Fortran program called THDSN for steady state best estimate Thermal Hydraulic (TH) calculations of plate type fuel RRs. Beside calculating the fuel and coolant temperature distribution and pressure gradient in an average and hot channel the program calculates the safety limits and margins against the critical phenomena encountered in RR such as the burnout heat flux and the onset of flow instability. Well known TH correlations for calculating the safety parameters are used. THDSN program is verified by comparing its results for 2 and 10 MW benchmark reactors with that published in IAEA publications and good agreement is found. Also the program results are compared with those published for other programs such as PARET and TERMIC. An extension for this program is underway to cover the transient TH calculations

  8. The reactor engineer program: creating a new workforce

    International Nuclear Information System (INIS)

    Summers, R.

    1993-01-01

    As the number of nuclear engineering schools continues to shrink across the U.S., talented professional engineers for the nuclear energy community must increasingly be found elsewhere. To meet its needs, therefore, the Office of Nuclear Reactor Regulation (NRR) established an Intern Program to bring new talent into the NRC. The two-year program includes 17 weeks of technical training, and 4 or 5 rotational assignments, including at least 4 months at a commercial nuclear power plant site. The key to the success of the program is the full support of NRR high-level management

  9. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

    Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

    2003-01-01

    In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

  10. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1999-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  11. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  12. Expert system for control rod programming of boiling water reactors

    International Nuclear Information System (INIS)

    Fukuzaki, T.; Yoshida, K.; Kobayashi, Y.; Matsuura, H.; Hoshi, K.

    1986-01-01

    Control rod programming, one of the main tasks in reactor core management of boiling water reactors (BWRs), can be successfully accomplished by well-experienced engineers. By use of core performance evaluation codes, their knowledge plays the main role in searching through optimal control rod patterns and exposure points for adjusting notch positions and exchanging rod patterns. An expert system has been developed, based on a method of knowledge engineering, to lighten the engineer's load in control rod programming. This system utilizes an inference engine suited for planning/designing problems, and stores the knowledge of well-experienced engineers in its knowledge base. In this report, the inference engine, developed considering the characteristics of the control rod programming, is introduced. Then the constitution and function of the expert system are discussed

  13. Program for aerodynamic performance tests of helium gas compressor model of the gas turbine high temperature reactor (GTHTR300)

    International Nuclear Information System (INIS)

    Takada, Shoji; Takizuka, Takakazu; Kunimoto, Kazuhiko; Yan, Xing; Itaka, Hidehiko; Mori, Eiji

    2003-01-01

    Research and development program for helium gas compressor aerodynamics was planned for the power conversion system of the Gas Turbine High Temperature Reactor (GTHTR300). The axial compressor with polytropic efficiency of 90% and surge margin more than 30% was designed with 3-dimensional aerodynamic design. Performance and surge margin of the helium gas compressor tends to be lower due to the higher boss ratio which makes the tip clearance wide relative to the blade height, as well as due to a larger number of stages. The compressor was designed on the basis of methods and data for the aerodynamic design of industrial open-cycle gas-turbine. To validate the design of the helium gas compressor of the GTHTR300, aerodynamic performance tests were planned, and a 1/3-scale, 4-stage compressor model was designed. In the tests, the performance data of the helium gas compressor model will be acquired by using helium gas as a working fluid. The maximum design pressure at the model inlet is 0.88 MPa, which allows the Reynolds number to be sufficiently high. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  14. International topical meeting on research reactor fuel management (RRFM) - United States Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) acceptance program: 2007 update

    International Nuclear Information System (INIS)

    Messick, C.E.; Taylor, J.L.

    2007-01-01

    The Nuclear Weapons Non-proliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, adopted by The United States Department of Energy (DOE), in consultation with the Department of State in May 1996, has been extended to expire May 12, 2016, providing an additional 10 years to return fuel to the U.S. This paper provides a brief update on the program, now transferred to the National Nuclear Security Administration (NNSA), and discusses program initiatives and future activities. The goal of the program continues to be recovery of nuclear materials (27 countries have participated so far, returning a total of 7620 spent nuclear fuel elements), which could otherwise be used in weapons, while assisting other countries to enjoy the benefits of nuclear technology. More than ever before, DOE and reactor operators need to work together to schedule shipments as soon as possible, to optimize shipment efficiency over the remaining years of the program. The NNSA is seeking feedback from research reactor operators to help us understand ways to include eligible reactor who have not yet participated in the program

  15. Qualification program for JHR fuel elements: Irradiation of the first JHR test assembly in the BR2-Evita loop

    International Nuclear Information System (INIS)

    Anselmet, M.-C.; Lemoine, P.; Koonen, E.; Benoit, P.; Gouat, P.; Claes, W.; Geens, F.; Miras, G.; Brisson, S.

    2010-01-01

    An experimental program has been designed by CEA to qualify the behaviour of the JHR fuel under conditions representative of the reactor operating ones. This program uses the SCK.CEN facilities, irradiating JHR lead test elements in the BR2 reactor, inside its central channel which has been particularly arranged for this objective (Evita loop). As a first step in the program, a two cycle irradiation (4 weeks by cycle) started mid-July 2009 and ended mid-November (EVITA-1). After a cooling phase, this first JHR lead test element will be submitted to post-irradiation examination. The second JHR test element began its irradiation in the first quarter of 2010; its unloading is planned before the end of 2010, after 5 cycles in the BR2 reactor. The results of these two experiments are expected as input information for the Safety Authority Report. This paper presents the qualification program with the objectives assigned to each phase (irradiation, examination). A first interpretation of the irradiation data for the first element is presented, so as the information available on the progress of the following phases of the programme. (author)

  16. JOYO modification program for demonstration tests of FBR innovative technology development

    International Nuclear Information System (INIS)

    Yoshimi, H.; Hachiya, Y.

    1990-01-01

    A plan is under way at PNC to modify the experimental fast reactor JOYO. The project is called MARK-III (MK-III) program. The purpose of MK-III is to expand the function of JOYO, and to make it possible to receive demonstration tests of new or high level technologies for FBR development. The MK-III program consists of two main modifications: conversion to a highly efficient irradiation facility; and a modification for demonstration testing of new technologies and concepts that have a high potential to reduce FBR plant construction cost, to evaluate plant reliability and to improve plant safety. These modifications are scheduled to start in 1991

  17. A status report on the integral fast reactor fuels and safety program

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Seidel, B.R.

    1990-01-01

    The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor (ALMR) concept being developed at Argonne National Laboratory. The IFR program is specifically responsible for the irradiation performance, advanced core design, safety analysis, and development of the fuel cycle for the US Department of Energy's ALMR program. The basic elements of the IFR concept are (a) metallic fuel, (b) liquid-sodium cooling, (c) modular, pool-type reactor configuration, (d) an integral fuel cycle based upon pyrometallurgical processing. The most significant safety aspects of the IFR program result from its unique fuel design, a ternary alloy of uranium, plutonium, and zirconium. This fuel is based on experience gained through > 25 yr operation of the Experimental Breeder Reactor II (EBR-II) with a uranium alloy metallic fuel. The ultimate criteria for fuel pin design is the overall integrity at the target burnup. The probability of core meltdown is remote; however, a theoretical possibility of core meltdown remains. The next major step in the IFR development program will be a full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. The IFR fuel cycle closure based on pyroprocessing will also have a dramatic impact on waste management options and on actinide recycling

  18. Reactor technology. Progress report, January--March 1978

    International Nuclear Information System (INIS)

    Warren, J.L.

    1978-07-01

    Progress is reported in eight program areas. The nuclear Space Electric Power Supply Program examined safety questions in the aftermath of the COSMOS 954 incident, examined the use of thermoelectric converters, examined the neutronic effectiveness of various reflecting materials, examined ways of connecting heat pipes to one another, studied the consequences of the failure of one heat pipe in the reactor core, and did conceptual design work on heat radiators for various power supplies. The Heat Pipe Program reported progress in the design of ceramic heat pipes, new application of heat pipes to solar collectors, and final performance tests of two pipes for HEDL applications. Under the Nuclear Process Heat Program, work continues on computer codes to model a pebble bed high-temperature gas-cooled reactor, adaptation of a set of German reactor calculation codes to use on U.S. computers, and a parametric study of a certain resonance integral required in reactor studies. Under the Nonproliferation Alternative Sources Assessment Program LASL has undertaken an evaluation of a study of gaseous core reactors by Southern Science Applications, Inc. Independently LASL has developed a proposal for a comprehensive study of gaseous uranium-fueled reactor technology. The Plasma Core Reactor Program has concentrated on restacking the beryllium reflector and redesigning the nuclear control system. The status of and experiments on four critical assemblies, SKUA, Godiva IV, Big Ten, and Flattop, are reported. The Nuclear Criticality Safety Program carried out several tasks including conducting a course, doing several annual safety reviews and evaluating the safety of two Nevada test devices. During the quarter one of the groups involved in reactor technology has acquired responsibility for the operation of a Cockroft-Walton accelerator. The present report contains information on the use of machine and improvements being made in its operation

  19. Research and development program in reactor safety for NUCLEBRAS

    International Nuclear Information System (INIS)

    Pinheiro, R.B.; Resende Lobo, A.A. de; Horta, J.A.L.; Avelar Esteves, F. de; Lepecki, W.P.S.; Mohr, K.; Selvatici, E.

    1984-01-01

    With technical assistance from the IAEA, it was established recently an analytical and experimental Research and Development Program for NUCLEBRAS in the area of reactor safety. The main objectives of this program is to make possible, with low investments, the active participation of NUCLEBRAS in international PWR safety research. The analytical and experimental activities of the program are described with some detail, and the main results achieved up to now are presented. (Author) [pt

  20. Requirements, needs, and concepts for a new broad-application test reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Fletcher, C.D.; Denison, A.B.; Liebenthal, J.L.

    1992-01-01

    For a variety of reasons, including (a) the increasing demands of the 1990s regulatory environment, (b) limited existing test capactiy and capability to satisfy projected future testing missions, and (c) an expected increasing need for nuclear information to support development of advanced reactors, there is a need for requirements and preliminary concepts for a new broad-application test reactor (BATR). These requirements must include consideration not only for a broad range of projected testing missions but also for current and projected regulatory compliance and safety requirements. The requirements will form the basis for development and assessment of preconceptual reactor designs and lead to the identification of key technologies to support the government's long-term strategic and programmatic planning. This paper outlines the need for a new BATR and suggests a few preliminary reactor concepts that can meet that need