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Sample records for reactor characterization program

  1. 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)

  2. A risk characterization of safety research areas for integral fast reactor program planning

    International Nuclear Information System (INIS)

    Mueller, C.J.; Cahalan, J.E.; Hill, D.J.; Kramer, J.M.; Marchaterre, J.F.; Pedersen, D.R.; Sevy, R.H.; Tibbrook, R.W.; Wei, T.Y.; Wright, A.E.

    1988-01-01

    This paper characterizes the areas of integral fast reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure to critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR safety and related base technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorities

  3. A risk characterization of safety research areas for Integral Fast Reactor program planning

    International Nuclear Information System (INIS)

    Mueller, C.J.; Cahalan, J.E.; Hill, D.J.

    1988-01-01

    This paper characterizes the areas of Integral Fast Reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure of critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR Safety and related Base Technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorites

  4. 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

  5. Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

    International Nuclear Information System (INIS)

    Smith, Curtis; Rabiti, Cristian; Martineau, Richard; Szilard, Ronaldo

    2016-01-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly ''over-design'' portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as ''safety margin.'' Historically, specific safety margin provisions have been formulated, primarily based on ''engineering judgment.''

  6. Light Water Reactor Sustainability Program: Risk-Informed Safety Margins Characterization (RISMC) Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

  7. 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

  8. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. Extended burnup demonstration: reactor fuel program. Pre-irradiation characterization and summary of pre-program poolside examinations. Big Rock Point extended burnup fuel

    International Nuclear Information System (INIS)

    Exarhos, C.A.; Van Swam, L.F.; Wahlquist, F.P.

    1981-12-01

    This report is a resource document characterizing the 64 fuel rods being irradiated at the Big Rock Point reactor as part of the Extended Burnup Demonstration being sponsored jointly by the US Department of Energy, Consumers Power Company, Exxon Nuclear Company, and General Public Utilities. The program entails extending the exposure of standard BWR fuel to a discharge average of 38,000 MWD/MTU to demonstrate the feasibility of operating fuel of standard design to levels significantly above current limits. The fabrication characteristics of the Big Rock Point EBD fuel are presented along with measurement of rod length, rod diameter, pellet stack height, and fuel rod withdrawal force taken at poolside at burnups up to 26,200 MWD/MTU. A review of the fuel examination data indicates no performance characteristics which might restrict the continued irradiation of the fuel

  15. Light Water Reactor Sustainability Program Risk-Informed Safety Margins Characterization (RISMC) PathwayTechnical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Curtis Smith; Cristian Rabiti; Richard Martineau

    2012-11-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). As the current Light Water Reactor (LWR) NPPs age beyond 60 years, there are possibilities for increased frequency of Systems, Structures, and Components (SSCs) degradations or failures that initiate safety-significant events, reduce existing accident mitigation capabilities, or create new failure modes. Plant designers commonly “over-design” portions of NPPs and provide robustness in the form of redundant and diverse engineered safety features to ensure that, even in the case of well-beyond design basis scenarios, public health and safety will be protected with a very high degree of assurance. This form of defense-in-depth is a reasoned response to uncertainties and is often referred to generically as “safety margin.” Historically, specific safety margin provisions have been formulated, primarily based on “engineering judgment.”

  16. 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

  17. Breazeale Reactor Modernization Program

    International Nuclear Information System (INIS)

    Davison, C. C.

    2003-01-01

    The Penn State Breazeale Nuclear Reactor is the longest operating licensed research reactor in the nation. The facility has played a key role in educating scientists, engineers and in providing facilities and services to researchers in many different disciplines. In order to remain a viable and effective research and educational institution, a multi-phase modernization project was proposed. Phase I was the replacement of the 25-year old reactor control and safety system along with associated wiring and hardware. This phase was fully funded by non-federal funds. Tasks identified in Phases II-V expand upon and complement the work done in Phase I to strategically implement state-of-the-art technologies focusing on identified national needs and priorities of the future

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  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. 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

  13. Reactor Containment Spray Technology Program

    Energy Technology Data Exchange (ETDEWEB)

    Row, T. H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    1968-12-15

    The design basis accident in water moderated power reactors is a loss-of-coolant accident in which water sprays are generally employed to control the containment pressure transient by condensing the released steam-air mixture. Additives to the spray have been proposed as a way to increase their usefulness by enhancing the removal of various forms of radioiodine from the containment atmosphere. A program to investigate the gas-liquid systems involved is co-ordinated by ORNL for the US Atomic Energy Commission. A basic part of the program is the search for various chemical additives that will increase the spray affinity for molecular iodine and methyl iodide. A method for evaluating additives was developed that measures equilibrium distribution coefficients for iodine between air and aqueous solutions. Additives selected are used in single drop-wind tunnel experiments where the circulating gas contains iodine or CH{sub 3}I. Mass transfer coefficients and transient distribution coefficients have been determined as a function of relative humidity, temperature, drop size, and solution pH and concentration. Tests have shown that surfactants and organic amines increase the solution ability to getter CH{sub 3}l. Results from single drop tests help in planning spray experiments in the Nuclear Safety Pilot Plant, a large ({approx}38 m{sup 3}) facility, where accident conditions are closely simulated. Iodine and CH{sub 3}I removal rates have been determined for a number of solutions, including 1 wt% Na{sub 2}S{sub 2}O{sub 3} + 3000 ppm B + 0.153 M NaOH and 3000 ppm B + 0.153 M NaOH. The additive has very little effect in removal of I{sub 2} with half-lives of less than 1 mm typical for any aqueous solution. These same solutions remove CH{sub 3}I with a half-life of one hour. Analytical models for the removal processes have been developed. Consideration is also being given to corrosion, thermal and radiation stability of the solutions. Radiation studies have indicated the loss

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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.)

  19. LOFT fuel modules design, characterization, and fabrication program

    International Nuclear Information System (INIS)

    Russell, M.L.

    1977-06-01

    The loss-of-fluid test [LOFT) fuel modules have evolved from a comprehensive five-year design, characterization, and fabrication program which has resulted in the accomplishment of many technical activities of interest in pressurized water reactor fuel design development and safety research. Information is presented concerning: determination of fundamental high-temperature reactor material properties; design invention related to in-core instrumentation attachment; implementation of advanced and/or unique fuel bundle characterization techniques; implementation of improved fuel bundle fabrication techniques; and planning and execution of a multimillion dollar design, characterization, and fabrication program for pressurized water reactor fuel

  20. 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

  1. 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.

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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)

  7. Characterization of reactor coolant by XRF

    Energy Technology Data Exchange (ETDEWEB)

    Legreid, G.; Beverskog, B. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  8. Characterization of reactor coolant by XRF

    International Nuclear Information System (INIS)

    Legreid, G.; Beverskog, B.

    2002-01-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  9. 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

  10. Uncertainty Characterization of Reactor Vessel Fracture Toughness

    International Nuclear Information System (INIS)

    Li, Fei; Modarres, Mohammad

    2002-01-01

    To perform fracture mechanics analysis of reactor vessel, fracture toughness (K Ic ) at various temperatures would be necessary. In a best estimate approach, K Ic uncertainties resulting from both lack of sufficient knowledge and randomness in some of the variables of K Ic must be characterized. Although it may be argued that there is only one type of uncertainty, which is lack of perfect knowledge about the subject under study, as a matter of practice K Ic uncertainties can be divided into two types: aleatory and epistemic. Aleatory uncertainty is related to uncertainty that is very difficult to reduce, if not impossible; epistemic uncertainty, on the other hand, can be practically reduced. Distinction between aleatory and epistemic uncertainties facilitates decision-making under uncertainty and allows for proper propagation of uncertainties in the computation process. Typically, epistemic uncertainties representing, for example, parameters of a model are sampled (to generate a 'snapshot', single-value of the parameters), but the totality of aleatory uncertainties is carried through the calculation as available. In this paper a description of an approach to account for these two types of uncertainties associated with K Ic has been provided. (authors)

  11. 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

  12. Manufacture of components for Canadian reactor programs

    International Nuclear Information System (INIS)

    Perry, L.P.

    Design features, especially those relating to calandrias, are pointed out for many CANDU-type reactors and the Taiwan research reactor. The special requirements shouldered by the Canadian suppliers of heavy reactor components are analyzed. (E.C.B.)

  13. 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.

  14. 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

  15. Characterization of liquid metal reactor materials

    International Nuclear Information System (INIS)

    Kuk, I. H.; Ryu, W. S.; Kim, H. H. and others

    1999-03-01

    The objectives of this report were to assess the material requirements for LMR environment, to select the optimum candidates for KALIMER components, to characterize the performance for establishing a database of the structural materials for KALIMER, and to develop the basic material technologies for the localization of the advanced materials. Stainless steel ingots were melted by VIM and hot-rolled to plate with the thickness of 15mm. The plate was solution-treated for 1 hr at 1100 deg C and then water-quenched. Specimens were taken parallel to the rolling direction of the plate. The effects of nitrogen and phosphorus were analyzed on the high temperature mechanical properties of 316MRP (Liquid Metal Reactor, Primary candidate material) stainless steels with the different nitrogen content from 0.04 to 0.15% and with the different phosphorus content from 0.002 to 0.02%. Heat treatment was performed to investigate the changes in microstructure and mechanical properties of Cr-Mo steels for LMR heat transfer tube materials and core materials. The Cr-Mo steels were normalized at the temperatures between 900 deg C and 1200 deg C for 1hrs and tempered at the temperatures between 500 deg C and 800 deg C for 2hrs. Conventional optical and electron micrographic studies were carried out to investigate the martensite lath structure, carbide indentification and carbide shape. Vickers microhardness was measured at room temperature using 10g load. Tensile properties were tested at high temperature. Charpy V-notch impact tests were also carried out at temperature between -120 deg C and +180 deg C. (author). 72 refs., 28 tabs., 244 figs

  16. Characterization of gamma field in the JSI TRIGA reactor

    Science.gov (United States)

    Ambrožič, Klemen; Radulović, Vladimir; Snoj, Luka; Gruel, Adrien; Guillou, Mael Le; Blaise, Patrick; Destouches, Christophe; Barbot, Loïc

    2018-01-01

    Research reactors such as the "Jožzef Stefan" Institute TRIGA reactor have primarily been designed for experimentation and sample irradiation with neutrons. However recent developments in incorporating additional instrumentation for nuclear power plant support and with novel high flux material testing reactor designs, γ field characterization has become of great interest for the characterization of the changes in operational parameters of electronic devices and for the evaluation of γ heating of MTR's structural materials in a representative reactor Γ spectrum. In this paper, we present ongoing work on γ field characterization both experimentally, by performing γ field measurements, and by simulations, using Monte Carlo particle transport codes in conjunction with R2S methodology for delayed γ field characterization.

  17. A Study on the Flow Characterization in the Reactor Cavity

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jung; Ko, Kwang Jeok; Kim, Sung Hwan; Kim, Min Gyu; Cho, Yeon Ho; Kim, Hyun Min [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

    2016-10-15

    In this study, the flow characterization of the cooling air in reactor cavity nearby RCPSA has been analyzed by using a 3 dimensional model and the ANSYS CFX software in order to predict the Convective Heat Transfer Coefficient (CHTC) of the RCPSA. The Reactor Cavity is the annular space by the concrete structure, the Reactor Cavity Pool Seal Assembly (RCPSA), which consists of the welded steel and is designed to be installed between the RV and the refueling pool floor, and the Reactor Vessel (RV). For such reason, the RCPSA should be designed to provide the cooling air passage for ventilation to circulate high temperature air passing by the RV during the reactor operation. It means that the RCPSA is influenced by the convection of cooling air and the thermal expansion of the RV. Therefore, the flow characterization at the reactor cavity is one of the factors of the RCPSA design during the reactor operation. The flow distribution of the cooling air in reactor cavity nearby RCPSA has been analyzed using ANSYS CFX software to obtain the CHTC at surface of the RCPSA. 1) The temperature from the RV and the insulation is one of the critical factors for the thermal gradient of the cooling air and the CHTC in the reactor cavity. 2) The rapid change of the CHTC in inner region nearby inner and outer flexure is related to the geometry shape of the RCPSA and velocity of cooling air.

  18. Programming Guidelines for FBD Programs in Reactor Protection System Software

    International Nuclear Information System (INIS)

    Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom; Lee, Jang Su

    2014-01-01

    Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows

  19. Programming Guidelines for FBD Programs in Reactor Protection System Software

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom [Division of Computer Science and Engineering College of Information and Communication, Konkuk University, Seoul (Korea, Republic of); Lee, Jang Su [Man-Machine Interface System team Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows.

  20. Research program plan: reactor vessels. Volume 1

    International Nuclear Information System (INIS)

    Vagins, M.; Taboada, A.

    1985-07-01

    The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

  1. 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

  2. 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

  3. Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

    International Nuclear Information System (INIS)

    Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

    1991-10-01

    This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs

  4. 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.)

  5. The RERTR demonstration experiments program at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K; King, J S [Department of Nuclear Engineering, University of Michigan (United States)

    1983-08-01

    The purpose of this paper is to highlight a major part of the experimental work which is being carried out at the Ford Nuclear Reactor (FNR) in conjunction with the RERTR program. A demonstration experiments program has been developed to: 1) characterize the FNR in sufficient detail to discern and quantify neutronic differences between the high and low enriched cores; 2) provide the theoretical group with measurements to benchmark their calculations. As with any experimental program associated with a reactor, stringent constraints limit the experiments which can be performed. Some experiments are performed routinely on the FNR (such as control rod calibrations), and much data is already available. Unfortunately, the accuracy we demand precludes using much of this earlier data. And in many cases, the requirement of precise (and copious) data has led to either developing new techniques (as in the case of rhodium mapping and neutron diffraction) or to further refinements on existing methods (as in the case of spectral unfolding). Nevertheless, we have tried to stay within the realm of recognized, well-established experimental methods in order to assuage any doubts about measured differences between HEU and LEU core parameters. This paper describes the principal results of the experiments performed so far.

  6. 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)

  7. Wide band characterization of wind turbine reactors

    DEFF Research Database (Denmark)

    Holdyk, Andrzej; Arana Aristi, Ivan; Holbøll, Joachim

    2013-01-01

    This paper presents the results of field measure-ments of the impedance of two commercial available wind turbine reactors, performed by a sweep frequency response analyzer, sFRA. The measurements were taken in the frequency range from 20Hz to 20MHz. Comparable frequency behavior was found in all ...

  8. 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.

  9. 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.

  10. 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.

  11. 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)

  12. 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

  13. 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.)

  14. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Busby, Jeremy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barnard, Cathy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-01

    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.

  15. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    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.

  16. The national standards program for research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1977-01-01

    In 1970 a standards committee called ANS-15 was established by the American Nuclear Society (ANS) to prepare appropriate standards for research reactors. In addition, ANS acts as Secretariat for a national standards committee N17 which is responsible to the American National Standards Institute (ANSI) for the national consensus efforts for standards related to research reactors. To date ANS-15 has completed or is working on 14 standards covering all aspects of the operation of research reactors. Of the 11 research reactor standards submitted to the ANSI N17 Committee since its inception, six have been issued as National standards, and the remaining are still in the process of review. (author)

  17. Light Water Reactor Sustainability Program: Integrated Program Plan

    International Nuclear Information System (INIS)

    2016-02-01

    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. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (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 industry to implement technology to exceed the performance of the current labor-intensive business model.

  18. Light Water Reactor Sustainability Program: Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2016-02-15

    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. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (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 industry to implement technology to exceed the performance of the current labor-intensive business model.

  19. 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

  20. 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.)

  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. Mechanical characterization of superalloys for space reactors

    International Nuclear Information System (INIS)

    Duchesne, J.

    1989-01-01

    The aim of this work is the selection of structural materials that can be used in the temperature range 600-900 0 C for a gas cooled space reactor producing electricity. Superalloys fit best the temperature range required. Five nickel base alloys are chosen for their good mechanical behaviour: HAYNES 230, HASTELLOY S, HASTELLOY X, HASTELLOY XR and PYRAD 38D. Metallography, tensile and hardness tests are realized. Sample contraction is evidenced for some creep tests, under low stress: 20MPa at 800 0 C, on HAYNES 230 and HASTELLOY X, probably related to the structural evolution of these materials corresponding to a decrease of the crystal parameter [fr

  3. 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

  4. Light Water Reactor Sustainability Program: Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-05-01

    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. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (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 industry to implement technology to exceed the performance of the current labor-intensive business model.

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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)

  11. Overview of the fast reactors fuels program

    International Nuclear Information System (INIS)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

  12. 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

  13. Mechanical characterization of superalloys for space reactors

    International Nuclear Information System (INIS)

    Duchesne, J.

    1989-01-01

    The purpose of this work is the choice of materials usable between 600 and 900 0 C for nuclear space reactor structures. The main criterion of selection for these materials is their good creep behaviour. Consequently, macroscopic theories of creep and several extrapolation methods were described. Superalloys seem the best materials for the studied range of temperatures. Five of them, base nickel, ones unusual in nuclear industry were selected for their good mechanical properties. Three of them are industrial alloys: the first, HAYNES 230 is a recent one, HASTELLOY S and X are more standard materials. The last two, HASTELLOY XR and PYRAD 38 D are issued from special fabrications. Creep tests metallographic investigations, hardness and tensile tests were performed. A contraction of samples was observed during some creep tests under a low stress, 20MPa at 800 0 C, for HAYNES 230 and HASTELLOY X. This could be due to a structural evolution of these materials connected to a decrease of the cristalline parameter. In addition, correlations were observed between certain characteristics determined from slow tensile tests and short duration creep tests. These correlations present a large interest because, at the present time, creep tests cannot be executed on irradiated materials in our laboratories. Consequently creep behaviour of irradiated materials seem may be deduced. Further studies are needed to explain and confirm the behaviour of the most interesting materials under low stresses: HAYNES 230 and HASTELLOY XR to anticipate their behaviour in working conditions [fr

  14. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

  15. Characterization of radioactive waste from nuclear power reactors

    International Nuclear Information System (INIS)

    Piumetti, Elsa H.; Medici, Marcela A.

    2007-01-01

    Different kinds of radioactive waste are generated as result of the operation of nuclear power reactors and in all cases the activity concentration of several radionuclides had to be determined in order to optimize resources, particularly when dealing with final disposal or long-term storage. This paper describes the three basic approaches usually employed for characterizing nuclear power reactor wastes, namely the direct methods, the semi-empirical methods and the analytical methods. For some radionuclides or kind of waste, the more suitable method or combination of methods applicable is indicated, stressing that these methods shall be developed and applied during the waste generation step, i.e. during the operation of the reactor. In addition, after remarking the long time span expected from waste generation to their final disposal, the importance of an appropriate record system is pointed out and some basic requirements that should be fulfilled for such system are presented. It is concluded that the tools for a proper characterization of nuclear reactor radioactive waste are available though such tools should be tailored to each specific reactor and their history. (author) [es

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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.)

  1. 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)

  2. 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

  3. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    International Nuclear Information System (INIS)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste

  4. 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

  5. Westinghouse Small Modular Reactor (SMR) Programe

    International Nuclear Information System (INIS)

    Shulyak, Nick

    2014-01-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) in which all primarycomponents associated with the nuclear steam supply system, including the steam generator and the pressurizer, are housed within the reactor vessel. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. This paper describes the design and functionality of the Westinghouse SMR, the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design drivers include safety, economics, reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR safety system design is passive, is based largely on the passive safety systems used in the AP1000 reactor, and provides mitigation of all design basis accidents without the need for offsite AC electrical power for a period of seven days. The economics of the Westinghouse SMR challenges the established approach of large Light Water Reactors (LWR) that utilized the economies of scale to reach economic competiveness. To serve the market expectation of smaller capital investment and cost competitive energy, a modular design approach is implemented within the Westinghouse SMR. The Westinghouse SMR building layout integrates the three basic design constraints of modularization; transportation, handling and module-joining technology. The integral Westinghouse SMR design eliminates large loop piping, which significantly reduces the flow area of postulated loss of coolant accidents (LOCAs). The Westinghouse SMR containment is a high

  6. 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

  7. U.S. Department of Energy Program of International Technical Cooperation for Research Reactor Utilization

    International Nuclear Information System (INIS)

    Chong, D.; Manning, M.; Ellis, R.; Apt, K.; Flaim, S.; Sylvester, K.

    2004-01-01

    The U.S. Department of Energy, National Nuclear Security Administration (DOE/NNSA) has initiated collaborations with the national nuclear authorities of Egypt, Peru, and Romania for the purpose of advancing the commercial potential and utilization of their respective research reactors. Under its Office of International Safeguards ''Sister Laboratory'' program, DOE/NNSA has undertaken numerous technical collaborations over the past decade intended to promote peaceful applications of nuclear technology. Among these has been technical assistance in research reactor applications, such as neutron activation analysis, nuclear analysis, reactor physics, and medical radioisotope production. The current collaborations are intended to provide the subject countries with a methodology for greater commercialization of research reactor products and services. Our primary goal is the transfer of knowledge, both in administrative and technical issues, needed for the establishment of an effective business plan and utilization strategy for the continued operation of the countries' research reactors. Technical consultation, cooperation, and the information transfer provided are related to: identification, evaluation, and assessment of current research reactor capabilities for products and services; identification of opportunities for technical upgrades for new or expanded products and services; advice and consultation on research reactor upgrades and technical modifications; characterization of markets for reactor products and services; identification of competition and estimation of potential for market penetration; integration of technical constraints; estimation of cash flow streams; and case studies

  8. Updated neutron spectrum characterization of SNL baseline reactor environments

    International Nuclear Information System (INIS)

    Griffin, P.J.; Kelly, J.G.; Vehar, D.W.

    1994-04-01

    This document provides SAND-II and MANIPULATE output listings from calculations used to derive the new spectrum-averaged integral parameters that were reported in volume 1. When used in conjunction with volume 1, this document provides an audit trail for the neutron radiation field characterization and supports current quality assurance initiatives. This document provides detailed information on the neutron spectrum characteristics of the primary Sandia National Laboratories' (SNL) reactor environments. The information in this volume is not intended for the casual user of the SNL reactor facilities. This detailed characterization of the neutron and gamma environments at the Sandia Pulsed Reactor (SPR) and the Annular Core Research Reactor (ACRR) is provided to aid the users who wish to convert the information given in the Radiation Metrology Laboratory (RML) dosimetry reports into other (non-silicon) measures of neutron damage. The spectra provided in these appendices can be used as a source term for Monte Carlo radiation transport calculations to study the impact of experimenter's test package on the neutron environment

  9. Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States); Peko, D. [US Dept. of Energy, Washington, DC (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Humrickhouse, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-06-01

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

  10. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

    2015-06-01

    In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary

  11. 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)

  12. 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)

  13. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  14. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    International Nuclear Information System (INIS)

    1979-12-01

    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials

  15. 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)

  16. 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

  17. 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)

  18. 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)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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

  5. Pacific Northwest Laboratory Monthly Activities Report for August 1966 AEC Division of Reactor Development and Technology Programs

    Energy Technology Data Exchange (ETDEWEB)

    SL Fawcett

    1966-08-01

    This report has the following sections: Summary; Civilian Power Reactors; Applied and Reactor Physics; Reactor Fuels and Materials; Engineering Development; Plutonium Recycle Program; and Nuclear Safety.

  6. 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)

  7. 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

  8. 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

  9. Integral Fast Reactor Program annual progress report, FY 1991

    International Nuclear Information System (INIS)

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. 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

  10. Integral Fast Reactor Program. Annual progress report, FY 1992

    Energy Technology Data Exchange (ETDEWEB)

    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&D.

  11. Integral Fast Reactor Program annual progress report, FY 1994

    International Nuclear Information System (INIS)

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

    1994-12-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R ampersand D

  12. Integral Fast Reactor Program. Annual progress report, FY 1993

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. 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 and D.

  13. Integral Fast Reactor Program. Annual progress report, FY 1993

    International Nuclear Information System (INIS)

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

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. 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 and D

  14. Japanese program of materials research for fusion reactors

    International Nuclear Information System (INIS)

    Hasiguti, R.R.

    1982-01-01

    The Japanese program of materials research for fusion reactors is described based on the report to the Nuclear Fusion Council, the project research program of the Ministry of Education, Science and Culture, and other official documents. The alloy development for the first wall and its radiation damage are the main topics discussed in this paper. Materials viewpoints for the Japanese Tokamak facilities and the problems of irradiation facilities are also discussed. (orig.)

  15. Characterization of graphite-matrix pulsed reactor fuels

    International Nuclear Information System (INIS)

    Karnes, C.H.; Marion, R.H.

    1976-01-01

    The performance of the Annular Core Pulsed Reactor (ACPR) is being upgraded in order to accommodate higher fluence experiments for fast reactor fuel element transient and safety studies. The increased fluence requires a two-zone core with the inner zone containing fuel having a high enthalpy and the capability of withstanding very high temperatures during both pulsed and steady state operation. Because the fuel is subjected to a temperature risetime of 2 to 5 ms and to a large temperature difference across the diameter, fracture due to thermal stresses is the primary failure mode. One of the fuels considered for the high enthalpy inner region is a graphite-matrix fuel containing a dispersion of uranium--zirconium carbide solid solution particles. A program was initiated to optimize the development of this class of fuel. This summary presents results on formulations of fuel which have been fabricated by the Materials Technology Group of the Los Alamos Scientific Laboratory

  16. 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

  17. 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)

  18. Fractal characterization for noise signal validation in power reactors

    International Nuclear Information System (INIS)

    Aguilar Martinez, Omar

    2003-01-01

    Up to now, a great variety of methods is used for the dynamical characterization of different components of Nuclear Power Plants (NPPs). With this aim, time and spectral analysis are usually considered, and different tools of non-stationary and non-gaussian analysis are also presented. When applying non-lineal dynamics theory for noise signal validation purposes in power reactors, the extraction of fractal echoes plays a main role. Fractal characterization for noise signal validation purposes can be integrated to the task of processing and acquisition of time signals in noise (fluctuation parameters) analysis systems. The possibility of discrimination between deterministic chaotic signals and pure noise signals has been incorporated, as a complement; to noise signals analysis in normal and anomalous operational conditions in NPPs using a fractal approach. In this work the detailed analysis of a neutronic sensor response is considered and the fractal characterization of its dynamics state (i.e. sensor line) for noise signal classification, it is presented. The experiment from where the time series (signals) were obtained, was carried out at the Research Reactor of the Technical University of Budapest, Hungary, during a model experiment for ageing process study of in-core neutron detectors (author)

  19. 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

  20. 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

  1. 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

  2. Status of Away From Reactor spent fuel storage program

    International Nuclear Information System (INIS)

    King, F.D.

    1979-07-01

    The Away From Reactor (AFR) Spent Fuel Program that the US Department of Energy established in 1977 is intended to preclude the shutting down of commercial nuclear power reactors because of lack of storage space for spent fuel. Legislation now being considered by Congress includes plans to provide storage space for commercial spent fuel beginning in 1983. Utilities are being encouraged to provide as much storage space as possible in their existing storage facilities, but projections indicate that a significant amount of AFR storage will be required. The government is evaluating the use of both existing and new storage facilities to solve this forecasted storage problem for commercial spent fuel

  3. 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

  4. 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

  5. Characterization of decommissioned reactor internals: Monte Carlo analysis technique

    International Nuclear Information System (INIS)

    Reid, B.D.; Love, E.F.; Luksic, A.T.

    1993-03-01

    This study discusses computer analysis techniques for determining activation levels of irradiated reactor component hardware to yield data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. The study recommends the Monte Carlo Neutron/Photon (MCNP) computer code as the best analysis tool for this application and compares the technique to direct sampling methodology. To implement the MCNP analysis, a computer model would be developed to reflect the geometry, material composition, and power history of an existing shutdown reactor. MCNP analysis would then be performed using the computer model, and the results would be validated by comparison to laboratory analysis results from samples taken from the shutdown reactor. The report estimates uncertainties for each step of the computational and laboratory analyses; the overall uncertainty of the MCNP results is projected to be ±35%. The primary source of uncertainty is identified as the material composition of the components, and research is suggested to address that uncertainty

  6. 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

  7. 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

  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. 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

  10. 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

  11. Materials Characterization Center program plan

    International Nuclear Information System (INIS)

    Nelson, R.D.; Ross, W.A.; Hill, O.F.; Mendel, J.E.; Merz, M.D.; Turcotte, R.P.

    1980-03-01

    The Materials Characterization Center (MCC) has been established at Pacific Northwest Laboratory as part of the Materials Characterization Organization for providing an authoritative, referenceable basis for establishing nuclear waste material properties and test methods. The MCC will provide a data base that will include information on the components of the waste emplacement package - the spent fuel or processed waste form and the engineered barriers - and their interaction with each other and as affected by the environment. The MCC will plan materials testing, develop and document procedures, collect and analyze existing materials data, and conduct tests as necessary

  12. 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

  13. The Canadian R and D program targeted at CANDU reactors

    International Nuclear Information System (INIS)

    Moeck, E.O.

    1988-01-01

    CANDU reactors produce electricity cheaply and reliably, with miniscule risk to the population and minimal impact on the environment. About half of Ontario's electricity and a third of New Brunswick's are generated by CANDU power plants. Hydro Quebec and utilities in Argentina, India, Pakistan, and the Republic of Korea also successfully operate CANDU reactors. Romania will soon join their ranks. The proven record of excellent performance of CANDUs is due in part to the first objective of the vigorous R and D program: namely, to sustain and improve existing CANDU power-plant technology. The second objective is to develop improved nuclear power plants that will remain competitive compared with alternative energy supplies. The third objective is to continue to improve our understanding of the processes underlying reactor safety and develop improved technology to mitigate the consequences of upset conditions. These three objectives are addressed by individual R and D programs in the areas of CANDU fuel channels, reduced operating costs, reduced capital costs, reactor safety research, and IAEA safeguards. The work is carried out mainly at three centres of Atomic Energy of Canada Limited--the Chalk River Nuclear Laboratories, the Whiteshell Nuclear Research Establishment, and the Sheridan Park Engineering Laboratories--and at Ontario Hydro's Research Laboratories. Canadian universities, consultants, manufacturers, and suppliers also provide expertise in their areas of specialization

  14. Microelectronics materials characterization studies at the Cornell TRIGA Reactor

    International Nuclear Information System (INIS)

    McGuire, Stephen C.

    1992-01-01

    The Cornell program of microelectronics materials characterization by neutron activation analysis (NAA) is described. Experimental details and results from the successful application of NAA to silicon germanium circuit structures and nickel silicide layers are presented. In doing so, the potential for using X rays from isotopes that decay by electron capture is demonstrated. (author)

  15. Computer code qualification program for the Advanced CANDU Reactor

    International Nuclear Information System (INIS)

    Popov, N.K.; Wren, D.J.; Snell, V.G.; White, A.J.; Boczar, P.G.

    2003-01-01

    Atomic Energy of Canada Ltd (AECL) has developed and implemented a Software Quality Assurance program (SQA) to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. This paper provides an overview of the computer programs used in Advanced CANDU Reactor (ACR) safety analysis, and assessment of their applicability in the safety analyses of the ACR design. An outline of the incremental validation program, and an overview of the experimental program in support of the code validation are also presented. An outline of the SQA program used to qualify these computer codes is also briefly presented. To provide context to the differences in the SQA with respect to current CANDUs, the paper also provides an overview of the ACR design features that have an impact on the computer code qualification. (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)

    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)

  17. 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)

  18. Characterization monitoring & sensor technology crosscutting program

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    The purpose of the Characterization, Monitoring, and Sensor Technology Crosscutting Program (CMST-CP) is to deliver appropriate characterization, monitoring, and sensor technology (CMST) to the OFfice of Waste Management (EM-30), the Office of Environmental Restoration (EM-40), and the Office of Facility Transition and Management (EM-60).

  19. 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

  20. Nuclear data needs for US fast reactor programs

    International Nuclear Information System (INIS)

    Daughtry, J.W.; Rawlins, J.A.

    1985-05-01

    Recent developments in US Fast Reactor Programs are reviewed to provide a background for the nuclear data needs of these programs. Innovative designs, inherent safety, and space nuclear power are receiving increased emphasis. Longstanding and newly-identified nuclear data requirements are reviewed. These requirements are based on information obtained early in 1985 in response to an inquiry sent out by the authors. Finally, plans are outlined for development of an adjusted cross section set for FFTF reload design. The adjustment process suggests some possible changes in ENDF/B-V nuclear data

  1. Preventive maintenance program for a research and production reactor

    International Nuclear Information System (INIS)

    Rico, N.A.

    1990-01-01

    This program proposes a simple, rapid and efficient methodology for the task of developing a really preventive maintenance discipline. Moreover, the lower cost of its application -since it must satisfy the plant's budget-. To this purpose, an extremely economical and easily obtainable infrastructure is proposed. The following stage is referred to the commissioning system, subsequent supervision and follow-up. The experience gained from the two reactors as RA-6 (Bariloche Atomic Center) and NUR (RAE) of Argelia. Finally, the interacting characteristic of this program, since it may be rapidly adapted to different dimensions of plants, laboratories, etc., must be pointed out. (Author) [es

  2. New initiatives in the U.S. Reactor Inspection Program

    International Nuclear Information System (INIS)

    Volgenau, Ernst.

    1977-01-01

    Recently, the United States Nuclear Regulatory Commission (NRC) has initiated a revised inspection approach that will involve placing inspectors full time onsite at all reactor sites. These resident inspectors will be supplemented by a performance appraisal inspection program that will incorporate thorough critical reviews of licensee facilities and an increased program of specific technical measurements to independently verify the accuracy and completeness of licensee work. To complement the inspection initiatives, the NRC is examining ways to expand its enforcement sanctions and to motivate safe licensee performance. (Auth.) [fr

  3. Characterization of a capillary plasma reactor for carbon dioxide decomposition

    International Nuclear Information System (INIS)

    Mori, Shinsuke; Yamamoto, Aguru; Suzuki, Masaaki

    2006-01-01

    The decomposition of carbon dioxide in a plasma reactor was investigated experimentally, using capillary discharge tubes with a diameter of 0.5 or 3.0 mm and a length of 25, 50, 75, 100 or 150 mm. The chemical composition of the reaction products and the current-voltage characteristics were measured over a pressure range of 3.33-120 Torr, and the CO 2 conversion rates and reduced electric fields were calculated. The results show that the influence of downscaling on the reduced electric fields can be well evaluated by adjusting both the current density, i, and the products of the pressure and the tube diameter, pd. However, the characteristics of CO 2 decomposition cannot be determined based on i and pd; they are better characterized by i and p. It can be deduced from our experimental results that the CO 2 conversion rate is predominated by the electron impact CO 2 dissociation and gas phase reverse reactions even in a capillary plasma reactor

  4. Minutes of the 14th Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) meeting, October 1-5, 1984

    International Nuclear Information System (INIS)

    1984-01-01

    Topics discussed include: ASTM LWR standards; trend curves, PSF, and other test reactor metallurgical programs; PSF dosimetry and metallurgical capsule neutron and gamma 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

  5. Materials characterization for advanced pressurized water reactors: Pt. 2

    International Nuclear Information System (INIS)

    Little, E.A.; Gage, G.

    1994-01-01

    A compilation and overview is presented of the experimental techniques available for characterization of the microstructural changes induced by neutron irradiation of PWR pressure vessel steels, and directed towards monitoring of embrittlement processes by examination of surveillance samples from advanced reactor systems. The microstructural features of significance include copper precipitation, dislocation loop and/or microvoid matrix damage and grain boundary solute segregation. The techniques of transmission electron microscopy, field-emission gun scanning transmission electron microscopy, small angle neutron scattering, positron annihilation and field-ion microscopy have all developed to a degree of sophistication such that they are capable of providing detailed microstructural information in these areas, and afford considerable insight into embrittlement processes when used in combination. (author)

  6. Dosimetry characterization of the Godiva Reactor under burst conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hickman, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hudson, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wong, C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ward, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wilson, C. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Clark, L. [Atomic Weapons Establishment (AWE), Berkshire (United Kingdom); Trompier, F. [Inst. for Radiation Protection and Nuclear Safety, Fontenay-aux-Roses (France)

    2017-06-22

    A series of sixteen (16) burst irradiations were performed in May 2014, fifteen of which were part of an international collaboration to characterize the Godiva IV fast burst reactor at the National Criticality Experiments Research Center (NCERC). Godiva IV is a bare cylindrical assembly of approximately 65 kg of highly enriched uranium fuel (93.2% 235U metal alloyed with 1.5% molybdenum for strength) and is designed to perform controlled prompt critical excursions (Myers 2010, Goda 2013). Twelve of the irradiations were dedicated to neutron spectral measurements using a Bonner multiple sphere spectrometer. Three irradiations, with core temperature increases of 71.1°C, 136.9°C, and 229.9°C, were performed for generating comparative fluence data, establishing corrections for varying heights, testing linearity with burst temperature, and establishing gamma dose characteristics.

  7. 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)

  8. Planned Scientific programs around the Triga Mark 2 Reactor

    International Nuclear Information System (INIS)

    Majah, M Ibn.

    2007-01-01

    Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

  9. Characterization, Monitoring and Sensor Technology Integrated Program

    International Nuclear Information System (INIS)

    1993-01-01

    This booklet contains summary sheets that describe FY 1993 characterization, monitoring, and sensor technology (CMST) development projects. Currently, 32 projects are funded, 22 through the OTD Characterization, Monitoring, and Sensor Technology Integrated Program (CMST-IP), 8 through the OTD Program Research and Development Announcement (PRDA) activity managed by the Morgantown Energy Technology Center (METC), and 2 through Interagency Agreements (IAGs). This booklet is not inclusive of those CMST projects which are funded through Integrated Demonstrations (IDs) and other Integrated Programs (IPs). The projects are in six areas: Expedited Site Characterization; Contaminants in Soils and Groundwater; Geophysical and Hydrogeological Measurements; Mixed Wastes in Drums, Burial Grounds, and USTs; Remediation, D ampersand D, and Waste Process Monitoring; and Performance Specifications and Program Support. A task description, technology needs, accomplishments and technology transfer information is given for each project

  10. 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

  11. 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

  12. 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

  13. Mechanical, chemical and radiological characterization of the graphite of the UNGG reactors type

    International Nuclear Information System (INIS)

    Bresard, I.; Bonal, J.P.

    2000-01-01

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

  14. 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

  15. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

  16. The United States advanced light water reactor (USALWR) development program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Devine, J.C. Jr.; Sugnet, W.R.

    1987-01-01

    For the United States Nuclear Power industry to remain viable, it must be prepared to meet the expected need for a new generation capacity in the late 90s with an improved reactor system. The best hope of meeting this requirement is with revolutionary changes to current LWR systems through simplification and re-evaluation of safety and operational design margins. In addition, the grid characteristics and the difficulty in raising capital for large projects indicate the smaller light water reactors (600 MWe) may play an important role in the next generation. A cooperative and coordinated program between EPRI, U.S. DOE, the major architect engineers, nuclear steam supply vendors, and the NRC in the U.S. has been undertaken with four major goals in mind

  17. The United States Advanced Light Water reactor (USALWR) development program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Devine, Jr.J.C.; Sugnet, W.R.

    1987-01-01

    For the United States Nuclear power industry to remain viable, it must be prepared to meet the expected need for a new generation capacity in the late 90s with an improved reactor system. The best hope of meeting this requirement is with revolutionary changes to current LWR systems through simplification and re-evaluation of safety and operational design margins. In addition, the grid characteristics and the difficulty in raising capital for large projects indicate the smaller light water reactors (600 MWe) may play an important role in the next generation. A cooperative and coordinated program between EPRI, U.S. DOE, the major architect engineers, nuclear steam supply vendors, and the NRC in the U.S. has been undertaken with four major goals in mind. (author)

  18. The US Advanced Liquid-Metal Reactor Program

    International Nuclear Information System (INIS)

    Brolin, E.C.

    1992-01-01

    Based on National Energy Strategy projections, utilities will be required to substantially increase electric generating capacity over the next 40 yr to meet economic growth requirements and replace retiring capacity. Although aggressive conservation measures can save up to 85 GW(electric), ∼195 GW(electric) of additional generating capcity will still be needed by 2010. Assuming startup of new plants around 2000, US Department of Energy (DOE) analyses show that nuclear power can contribute 195 GW(electric) of capacity by 2030, or ∼20% of total electric generation. The DOE is involved in a number of strategies designed to revitalize the nuclear power industry and enable it to meet this projected need for additional capacity. Among these is an integrated overall strategy for advanced reactor development and high-level waste management. A high priority in pursuit of this strategy is the Advanced Liquid-Metal Reactor (ALMR) Program

  19. Japanese Fast Reactor Program for Homogeneous Actinide Recycling

    International Nuclear Information System (INIS)

    Ishikawa, Makoto; Nagata, Takashi; Kondo, Satoru

    2008-01-01

    In the present report, the homogeneous actinide recycling scenario of Fast Reactor (FR) Cycle Technology Development Project (FaCT) is summarized. First, the scenario of nuclear energy policy in Japan are briefly reviewed. Second, the basic plan of Japan to manage all minor actinide (MA) by recycling is summarized objectives of which are the efficiency increase of uranium resources, the environmental burden reduction, and the increase of nuclear non-proliferation potential. Third, recent results of reactor physics study related to MA-loaded FR cores are briefly described. Fourth, typical nuclear design of MA-loaded FR cores in the FaCT project and their main features are demonstrated with the feasibility to recycle all MA in the future FR equilibrium society. Finally, the research and development program to realize the MA recycling in Japan is introduced, including international cooperation projects. (authors)

  20. 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.

  1. 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)

  2. Characterization of reactor neutron environments at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Kelly, J.G.; Luera, T.F.; Griffin, P.J.; Vehar, D.W.

    1994-01-01

    To assure quality in the testing of electronic parts in neutron radiation environments, Sandia National Laboratories (SNL) has incorporated modern techniques and procedures, developed in the last two decades by the radiation effects community, into all of its experimental programs. Attention to the application of all of these methodologies, experiment designs, nuclear data, procedures and controls to the SNL radiation services has led to the much more accurate and reliable environment characterizations required to correlate the effects observed with the radiation delivered

  3. Draft reclamation program plan for site characterization

    International Nuclear Information System (INIS)

    1989-08-01

    As part of its obligations under the Nuclear Waste Policy Act, as amended, the US Department of Energy (DOE) has developed an environmental program that is to be implemented during site characterization at the Yucca Mountain site. This site is proposed for the location of the nation's first high-level radioactive waste repository. A program for the reclamation of areas disturbed by site characterization is part of the overall environmental program for that site. This Reclamation Program Plan (RPP) describes the reclamation policy of the DOE for the Yucca Mountain site and presents an overview of the reclamation program. The RPP also provides an overview of the reclamation needs relative to site characterization; a review of legislation and requirements pertinent to reclamation; and a review of previous commitments made by the DOE to certain types of reclamation activities. The objective of the DOE reclamation program at Yucca Mountain is to return land disturbed by site-characterization activities to a stable ecological state with a form and productivity similar to the predisturbance state. The DOE will take all reasonable and necessary steps to achieve this objective. 19 refs., 2 tabs

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  5. 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)

  6. Programs with societal benefits at the Cornell University TRIGA reactor

    International Nuclear Information System (INIS)

    Clark, D.D.; Aderhold, H.C.; Hossain, T.Z.

    1993-01-01

    In its 30 yr of operation, the Cornell TRIGA reactor has been used for many educational and research programs that provide general benefits to society. In addition to supporting graduate-level education of nuclear scientists and engineers, it has been extensively used in undergraduate and graduate courses and research by nonspecialists and, through the medium of tours, in education of the general public. Some educational functions have been described previously. In this paper, examples are presented of research of societal interest in nonnuclear fields. The first two rely mainly on radiography, and the remaining five on neutron activation analysis (NAA)

  7. Severe accident sequence assessment for boiling water reactors: program overview

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1980-10-01

    The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

  8. The Nuclear Regulatory Commission's research reactor inspection program

    International Nuclear Information System (INIS)

    Constable, George L.

    1977-01-01

    The Division of Reactor Inspection Program's functional responsibilities are presented. The include (a) inspections and investigations necessary to determine whether licensees are complying with license provisions and rules, and to ascertain whether licensed operations are being conducted safely; (b) establishment of bases for the issuance or denial of a construction permit or license; (c) investigation of accidents, incidents, and theft or diversion of special nuclear materials; (d) enforcement actions; and (e) evaluation of licensed operations as a basis for recommending changes to standards and license conditions and for issuance of reports to the nuclear industry and the public

  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. 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)

  12. 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

  13. 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

  14. 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)

  15. Radiological characterization for small type light water reactor

    International Nuclear Information System (INIS)

    Tanaka, Ken-ichi; Ichige, Hideaki; Tanabe, Hidenori

    2011-01-01

    In order to plan a decommissioning, amount investigation of waste materials and residual radioactivity inventory evaluation must be performed at the first stage of preparatory tasks. These tasks are called radiological characterization. Reliable information from radiological characterization is crucial for specification of decommissioning plan. With the information, we can perform radiological safety analysis and optimize decommissioning scenario. Japan Atomic Power Company (JAPC) has already started preparatory tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1) that is the first commercial Small Type Light Water Reactor in Japan. To obtain reliable information about residual radioactivity inventory, we improved radioactivity inventory evaluation procedure. The procedure consists of neutron flux distribution calculation and radioactivity distribution calculation. We need a better understanding about characteristics of neutron transport phenomena in order to obtain reliable neutron flux distribution. Neutron flux was measured in Primary Containment Vessel (PCV) at 30 locations using activation foils. We chose locations where characteristic phenomena can be observed. Three dimensional (3D) neutron flux calculation was also performed to simulate continuous changes of neutron flux distribution. By assessing both the measured values and 3D calculation results, we could perform the calculation that simulates the phenomena well. We got knowledge about how to perform an appropriate neutron flux distribution calculation and also became able to calculate a reliable neutron flux distribution. Using the neutron flux distribution, we can estimate a reliable radioactivity distribution. We applied network-parallel-computing method to the estimation. And further we developed 'flux level approximation method' which use linear or parabola fitting method to estimation. Using these new methods, radioactivity by neutron irradiation, which is radioisotope formation, was calculated at

  16. Status of reduced enrichment program for research reactors in Japan

    International Nuclear Information System (INIS)

    Kaieda, Keisuke; Baba, Osamu; Nagaoka, Yoshiharu; Kanda, Keiji; Nakagome, Yoshihiro

    1999-01-01

    The reduced enrichment programs for the JRR-3M, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI) have been completed. The KUR of Kyoto University Research Reactor Institute (KURRI) has been partially completed and is still in progress under the Joint Study Program with Argonne National Laboratory (ANL). The JRR-3M commenced using LEU silicide fuel elements instead of LEU aluminide fuel elements in September, 1999. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and April 1994 the U.S. Government gave an approval to utilize HEU fuel in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until March 2004, then the full core conversion with LEU silicide will be done. The first shipment of spent fuels since 1974 was done in August, 1999. (author)

  17. 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

  18. Prioritization of R and D programs on probabilistic reactor safety

    International Nuclear Information System (INIS)

    Husseiny, A.A.

    1982-01-01

    An interactive computer code based on the multiattribute utility theory has been developed with graphic capabilities to use in selection of probabilistic reactor safety RandD programs. Utility values and proper graphic representation are made through lottery games on the computer terminal. The code is applied to prioritize a set of RandD programs on LWR safety based on attributes including regulatory issues, institutional issues and operation problems. The methodology is described here in detail with its applications. Some of the input includes statistical distributions and subjective judgments on institutional issues. The flexibility of the approach provides a tool for decision makers whether on individual or group level to assess LWR safety priorities and continuously update their strategies

  19. Cermet fuel for fast reactor – Fabrication and characterization

    Energy Technology Data Exchange (ETDEWEB)

    Mishra, Sudhir, E-mail: sudhir@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kutty, P.S.; Kutty, T.R.G. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Das, Shantanu [Uranium Extraction Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Dey, G.K. [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Kumar, Arun [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2013-11-15

    (U, Pu)O{sub 2} ceramic fuel is the well-established fuel for the fast reactors and (U, Pu, Zr) metallic fuel is the future fuel. Both the fuels have their own merits and demerits. Optimal solution may lie in opting for a fuel which combines the favorable features of both fuel systems. The choice may be the use of cermet fuel which can be either (U, PuO{sub 2}) or (Enriched U, UO{sub 2}). In the present study, attempt has been made to fabricate (Natural U, UO{sub 2}) cermet fuel by powder metallurgy route. Characterization of the fuel has been carried out using dilatometer, differential thermal analyzer, X-ray diffractometer, and Scanning Electron Microscope. The results show a high solidus temperature, high thermal expansion, presence of porosities, etc. in the fuel. The thermal conductivity of the fuel has also been measured. X-ray diffraction study on the fuel compact reveals presence of α U and UO{sub 2} phases in the matrix of the fuel.

  20. Program for studying fundamental interactions at the PIK reactor facilities

    International Nuclear Information System (INIS)

    Serebrov, A. P.; Vassiljev, A. V.; Varlamov, V. E.; Geltenbort, P.; Gridnev, K. A.; Dmitriev, S. P.; Dovator, N. A.; Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A.; Martemyanov, V. P.

    2016-01-01

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4′ channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4′ channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  1. The Program Planned for the Molten Salt Reactor Experiment

    International Nuclear Information System (INIS)

    Haubenreich, Paul N.

    1967-01-01

    This document outlines the program planned for the MSRE in fiscal years 1968 and 1969. It includes a bar diagram of the program, a critical-path type diagram of the operations, and a brief description of each task. In addition to the work at the reactor site, the outline also covers activities elsewhere at ORNL and by the AEC that directly affect the reactor schedule. The amount of detail and the accuracy with which we can estimate times varies considerably among the different items on the schedule. Some items, such as annual checkouts and core sample replacement, have been done before and our time estimates do not include any contingency, In the case of such tasks as planning, reviewing, and preparing for experiments or operations, we have set target dates that appear reasonable and we fully expect to meet these. Processing the salt is a different matter. If there are no unforeseen difficulties we should finish easily in the time shown, but the operation is in part a shakedown, so delays would not be too surprising, The time for modifying the system and adding fluoroborate is, of course, uncertain because the requirements are not yet known. As the requirements develop in more detail the estimate will be updated, but we do not foresee any major dislocation in the schedule, The scheduled time for preparation of enriching salt is becoming tight because of delays in facility construction. Should there be further delays in this key item, the entire schedule would have to be reconsidered.

  2. Program for studying fundamental interactions at the PIK reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Vassiljev, A. V.; Varlamov, V. E. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation); Geltenbort, P. [Institut Laue-Langevin (France); Gridnev, K. A. [St. Petersburg State University (Russian Federation); Dmitriev, S. P.; Dovator, N. A. [Russian Academy of Sciences, Ioffe Physical-Technical Institute (Russian Federation); Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P. [National Research Center Kurchatov Institute (Russian Federation); and others

    2016-05-15

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4′ channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4′ channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  3. Efficient characterization of fuel depletion in boiling water reactor

    International Nuclear Information System (INIS)

    Kim, S.H.

    1980-01-01

    An efficient fuel depletion method for boiling water reactor (BWR) fuel assemblies has been developed for fuel cycle analysis. A computer program HISTORY based on this method was designed to carry out accurate and rapid fuel burnup calculation for the fuel assembly. It has been usefully employed to study the depletion characteristics of the fuel assemblies for the preparation of nodal code input data and the fuel management study. The adequacy and the effectiveness of the assessment of this method used in HISTORY were demonstrated by comparing HISTORY results with more detailed CASMO results. The computing cost of HISTORY typically has been less than one dollar for the fuel assembly-level depletion calculations over the full life of the assembly, in contrast to more than $1000 for CASMO. By combining CASMO and HISTORY, a large number of expensive CASMO calculations can be replaced by inexpensive HISTORY. For the depletion calculations via CASMO/HISTORY, CASMO calculations are required only for the reference conditions and just at the beginning of life for other cases such as changes in void fraction, control rod condition and temperature. The simple and inexpensive HISTORY is sufficienty accurate and fast to be used in conjunction with CASMO for fuel cycle analysis and some BWR design calculations

  4. 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)

  5. 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

  6. 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

  7. Thermodynamic characterization of salt components for Molten Salt Reactor fuel

    NARCIS (Netherlands)

    Capelli, E.

    2016-01-01

    The Molten Salt Reactor (MSR) is a promising future nuclear fission reactor technology with excellent performance in terms of safety and reliability, sustainability, proliferation resistance and economics. For the design and safety assessment of this concept, it is extremely important to have a

  8. 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

  9. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  10. 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

  11. 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.

  12. Program status of the high temperature reactor development in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    1984-01-01

    The status of the HTR development program in the Federal Republic of Germany in 1984 is characterized by the beginning of a transition phase from a national program to a commercial program. In the last 20 years the HTR technology program was strongly, nearly completely supported by the Federal Government and the State Government of North-Rhine-Westfalia. Funding of the program up to now exceeded 5 billion DM. Within this framework it was possible to establish competent-reactor-system companies, to enable industries to supply HTR- specific components including fuel elements and nuclear graphites, to maintain the strong engagement of the national centre KFA Juelich in general R and D activities, to build and operate the AVR-plant for more than 16 years, to erect the demonstration plant THTR-300 now approaching completion and to build and operate many efficient test facilities. Thereby the HTR technology development achieved a stage of maturity which is not only considered to be most advanced, but is also ready now for commerical deployment. The assessment report which comprised both the fast breeder and the HTR development included all major impacts, such as history, status, prospects, benefits, industrial aspects and international developments of the technology. The program description is facilitated by distinguishing the five major program elements: AVR, THTR-300, THTR follow-up plant, nuclear process heat program, fuel cycle activities

  13. 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

  14. 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)

  15. 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

  16. Summary of detection, location, and characterization capabilities of AE for continuous monitoring of cracks in reactors

    International Nuclear Information System (INIS)

    Hutton, P.H.; Kurtz, R.J.; Friesel, M.A.; Pappas, R.A.; Skorpik, J.R.; Dawson, J.F.

    1984-10-01

    The objective of the program is to develop acoustic emission (AE) methods for continuous monitoring of reactor pressure boundaries to detect and evaluate crack growth. The approach involves three phases: develop relationships to identify crack growth AE signals and to use identified crack growth AE data to estimate flaw severity; evaluate and refine AE/flaw relationships through fatigue testing a heavy section vessel under simulated reactor conditions; and demonstrate continuous AE monitoring on a nuclear power reactor system

  17. MIT nuclear reactor laboratory high school teaching program

    International Nuclear Information System (INIS)

    Olmez, I.

    1991-01-01

    For the last 6 years, the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory's academic and scientific staff a have been conducting evening seminars for precollege science teachers, parents, and high school students from the New England area. These seminars, as outlined in this paper, are intended to give general information on nuclear technologies with specific emphasis on radiation physics, nuclear medicine, nuclear chemistry, and ongoing research activities at the MIT research reactor. The ultimate goal is to create interest or build on the already existing interest in science and technology by, for example, special student projects. Several small projects have already been completed ranging from environmental research to biological reactions with direct student involvement. Another outcome of these seminars was the change in attitudes of science teachers toward nuclear technology. Numerous letters have been received from the teachers and parents stating their previous lack of knowledge on the beneficial aspects of nuclear technologies and the subsequent inclusion of programs in their curriculum for educating students so that they may also develop a more positive attitude toward nuclear power

  18. Tank farm waste characterization Technology Program Plan

    International Nuclear Information System (INIS)

    Hohl, T.M.; Schull, K.E.; Bensky, M.S.; Sasaki, L.M.

    1989-03-01

    This document presents technological and analytical methods development activities required to characterize, process, and dispose of Hanford Site wastes stored in underground waste tanks in accordance with state and federal environmental regulations. The document also lists the need date, current (fiscal year 1989) funding, and estimate of future funding for each task. Also identified are the impact(s) if an activity is not completed. The document integrates these needs to minimize duplication of effort between the various programs involved

  19. Site characterization studies in the NWTS program

    International Nuclear Information System (INIS)

    Shipler, D.; Evans, G.

    1980-01-01

    The US Department of Energy (DOE) has the responsibility to identify sites and construct and operate facilities for the storage or isolation of spent fuel and/or reprocessing radioactive wastes from commercial nuclear power plants. The National Waste Terminal Storage (NWTS) Program has been initiated by the DOE to develop the technology and demonstrate the feasibility of burial and isolation of high level radioactive waste in deep geologic formations. The NTWS Program plan which sets forth the criteria, procedures, and other considerations required to characterize and select a site in a comprehensive stepwise manner is discussed. The plan is not specific to any given geologic medium but serves as a guide for site selection in any geohydrologic system deemed appropriate for consideration for a deep geologic repository. The plan will be used by all NWTS Project Offices in the conduct of their site characterization program. The plan will be updated, as warranted, to reflect technology development, National policies, rulemakings by regulatory agencies, and other changing political, social, and institutional considerations. Site characterization begins with the identification of regions believed to have suitable geologic, hydrologic, and environmental characteristics for repository siting. This is followed by an iterative process of data collection and analysis to identify areas and locations which appear most suitable for further investigations. In addition, screening studies of the DOE's nuclear complexes has led to the selection of the Nevada Test Site and the Hanford Site for further characterization studies. The site characterization process results in a number of candidate sites from which a site will be selected and proposed to the NRC for licensing

  20. TERRAPOWER, LLC TRAVELING WAVE REACTOR DEVELOPMENT PROGRAM OVERVIEW

    Directory of Open Access Journals (Sweden)

    PAVEL HEJZLAR

    2013-11-01

    burden. This paper describes the origins and current status of the TWR development program at TerraPower, LLC. Some of the areas covered include the key TWR design challenges and brief descriptions of TWR-Prototype (TWR-P reactor. Selected information on the TWR-P core designs are also provided in the areas of neutronic, thermal hydraulic and fuel performance. The TWR-P plant design is also described in such areas as; system design descriptions, mechanical design, and safety performance.

  1. Restructured site characterization program at Yucca Mountain

    International Nuclear Information System (INIS)

    Dyer, J.R.; Vawter, R.G.

    1995-01-01

    During 1994 and the early part of 1995, the US Department of Energy's Yucca Mountain Site Characterization Office (YMSCO) and its parent organization, the Office of Civilian Radioactive Waste Management (OCRWM) underwent a significant restructuring. Senior Department officials provided the leadership to reorient the management, technical, programmatic, and public interaction approach to the US High Level Radioactive Waste Disposal Program. The restructuring involved reorganizing the federal staff, conducting meaningful strategic planning, improving the management system, rationalizing contractor responsibilities, focusing upon major products, and increasing stakeholder involvement. The restructured program has prioritized technical and scientific activities toward meeting major regulatory milestones in a timely and cost-effective manner. This approach has raised concern among elements of technical, scientific, and oversight bodies that suitability and licensing decisions could be made without obtaining sufficient technical information for this first-of-its-kind endeavor. Other organizations, such as congressional committees, industrial groups, and rate payers believe characterization goals can be met in a timely manner and within the limitation of available funds. To balance these contrasting views in its decision making process, OCRWM management has made a special effort to communicate its strategy to oversight bodies, the scientific community and other stakeholders and to use external independent peer review as a key means of demonstrating scientific credibility. Site characterization of Yucca Mountain in Nevada is one of the key elements of the restructured program

  2. 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.

  3. 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).

  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. Advanced Light Water Reactor Program: Program management and staff review methodology

    International Nuclear Information System (INIS)

    Moran, D.H.

    1986-12-01

    This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond

  6. Status of reduced enrichment programs for research reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  7. Status of reduced enrichment programs for research reactors in Japan

    International Nuclear Information System (INIS)

    Kanda, Keiji; Nishihara, Hedeaki; Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo

    1997-01-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE

  8. GENP-2, Program System for Integral Reactor Perturbation

    International Nuclear Information System (INIS)

    Boioli, A.; Cecchini, G.P.

    1975-01-01

    1 - Description of problem or function: GENP-2 is a system of programs that use 'generalized perturbation theory' to calculate the perturbations of reactor integral characteristics which can be expressed by means of ratios between linear or bilinear functionals of the real and/or adjoint fluxes (e.g. reaction rate ratios), due to cross section perturbations. 2 - Method of solution: GENP-2 consists of the following codes: DDV, SORCI, CIAP-PMN and GLOBP-2D. DDV calculates the real or adjoint fluxes and power distribution using multigroup diffusion theory in 2-dimensions. SORCI uses the fluxes from DDV to calculate the real and/or adjoint general perturbation sources. CIAP-PMN reads the sources from SORCI and uses them in the real or adjoint generalised importance calculations (2 dimensions, multi- group diffusion). GLOBP-2D uses the importance calculated by CIAP-PMN, and the fluxes calculated by DDV, in generalised perturbation expressions to calculate the perturbation in the quantity of interest. 3 - Restrictions on the complexity of the problem: DDV although variably dimensioned has the following restrictions: - max. number of mesh points 6400; - max. number of mesh points in 1-dimension 81; - max. number of regions 6400; - max. number of energy groups 100; - if power distribution calculated, product of number of groups and number of regions 2500. The other programs have the same restrictions if applicable

  9. Thermodynamic characterization of the molten salt reactor fuel - 5233

    International Nuclear Information System (INIS)

    Capelli, E.; Konings, R.J.M.; Benes, O.

    2015-01-01

    The Molten Salt Reactor (MSR) has been selected as one of the Generation IV nuclear systems. The very unique feature of this reactor concept is the liquid nature of the fuel which offers numerous advantages concerning the reactor safety. Nowadays, the research in Europe is focused on an innovative concept, the MSFR (Molten Salt Fast Reactor), that combines the generic assets of molten salt as liquid fuel with those related to fast neutron reactors and the thorium fuel cycle. For the design and safety assessment of the MSFR concept, it is extremely important to have a thorough knowledge of the physico-chemical properties of fluorides salts, which is the class of materials that is the best suited for nuclear applications. Potential chemical systems have been critically reviewed and an extensive thermodynamic database describing the most relevant systems has been created at the Institute for Transuranium Elements of the Joint Research Centre (JRC). Thermochemical equilibrium calculations are a very important tool that allows the evaluation of the performance of several salt mixtures predicting their properties and thus the optimization of the fuel composition. The work combines the experimental determination of different salt properties with the modelling of the thermodynamic functions, using the Calphad method. An overview of the experimental work and the thermodynamic assessments will be given in this paper and different fuel options for the MSFR will be discussed. (authors)

  10. Fast reactor development program in France in 1997

    International Nuclear Information System (INIS)

    Langrand, J.C.; Hubert, G.; Marmonier, P.; Del Negro, R.; Saint-Arroman, G.; Coulon, P.; Mesnage, B.; Pillon, S.; Lefevre, J.C.

    1998-01-01

    , in spite of the availability of one and half core. This decision leads to drop the >. The Capra program was reoriented. Now, emphasis is put on burning the minor actinides (MA) and long-lived fusion products (LLFP), which answers the French 1991 law on radioactive wastes; as a consequence, the experimental programme proposed for Phenix is mainly devoted to this objective. This report presents also the status of Rapsodie, and the main achievements on the R and D programme on fast reactors. (author)

  11. Cadmium-emitter self-powered thermal neutron detector performance characterization & reactor power tracking capability experiments performed in ZED-2

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W., E-mail: physics@execulink.com [LaFontaine Consulting, Kitchener, Ontario (Canada); Zeller, M.B. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Nielsen, K. [Royal Military College of Canada, SLOWPOKE-2 Reactor, Kingston, Ontario (Canada)

    2014-07-01

    Cadmium-emitter self-powered thermal neutron flux detectors (SPDs), are typically used for flux monitoring and control applications in low temperature, test reactors such as the SLOWPOKE-2. A collaborative program between Atomic Energy of Canada, academia (Royal Military College of Canada (RMCC)) and industry (LaFontaine Consulting) was initiated to characterize the incore performance of a typical Cd-emitter SPD; and to obtain a definitive measure of the capability of the detector to track changes in reactor power in real time. Prior to starting the experiment proper, Chalk River Laboratories' ZED-2 was operated at low power (5 watts nominal) to verify the predicted moderator critical height. Test measurements were then performed with the vertical center of the SPD emitter positioned at the vertical mid-plane of the ZED-2 reactor core. Measurements were taken with the SPD located at lattice position L0 (near center), and repeated at lattice position P0 (in D{sub 2}O reflector). An ionization chamber (part of the ZED-2 control instrumentation) monitored reactor power at a position located on the south side of the outside wall of the reactor's calandria. These experiments facilitated measurement of the absolute thermal neutron sensitivity of the subject Cd-emitter SPD, and validated the power tracking capability of said SPD. Procedural details of the experiments, data, calculations and associated graphs, are presented and discussed. (author)

  12. Detection and characterization of flaws in segments of light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

    1988-01-01

    Studies have been conducted to determine flaw density in segments cut from light water reactor )LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (H SST) Program. Segments from the Hope Creek Unit 2 vessel and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication [with a through-wall dimension of ∼6 mm (∼0.24 in.)] was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications [i.e., for a total of approximately 6.8 m 2 (72 ft 2 ) of cladding surface]. (author)

  13. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  14. Neutron and gamma characterization within the FFTF reactor cavity

    International Nuclear Information System (INIS)

    Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

    1980-08-01

    Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables

  15. Characterization of four prestressed concrete reactor vessel liner steels

    International Nuclear Information System (INIS)

    Nanstad, R.K.

    1980-12-01

    A program of fracture toughness testing and analysis is being performed with PCRV steels for HTGRs. This report focuses on background information for the base materials and results of characterization testing, such as tensile and impact properties, chemical composition, and microstructural examination. The steels tested were an SA-508 class 1 forging, two plates of SA-537 class 1, and one plate of SA-537 class 2. Tensile requirements in effect at the time of procurement are met by all four steels. However, the SA-537 class 2 plate would not meet the minimum requirement for yield strength. Drop-weight and Charpy impact tests verified that the RT/sub NDT/ is equal to the NDT for each steel. Charpy impact energies at the NDT range from 40 J (30 ft-lb) for one heat of SA-537 class 1 to 100 J (74 ft-lb) for the SA-537 class 2 plate; upper-shelf energies range from 170 to 310 J (125 to 228 ft-lb) for the same two steels, respectively. The onset of upper-shelf energy occurred at temperatures ranging from 0 to 50 0 C

  16. 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

  17. Research and development program for PWR safety at the CEA reactor thermal hydraulics laboratories

    International Nuclear Information System (INIS)

    Bernard, M.

    1995-01-01

    Since the start of the French electronuclear program, the three partners Fermate, EDF and Cea (DRN and IPSN) have devoted considerable effort to research and development for safety issues. In particular an important program on thermal hydraulics was initiated at the beginning of the seventies. It is illustrated by the development of the CATHARE thermalhydraulic safety code which includes physical models derived from a large experimental support program and the construction of the BETHSY integral facility which is aimed to assess both the CATHARE code and the physical relevance of the accident management procedures to be applied on reactors. The state of the art on this program is described with particular emphasis on the capabilities and the assessment of the last version of CATHARE and the lessons drawn from 50 BETHSY tests performed so far. The future plans for safety research cover the following strategy: - to solve the few problems identified on present computing tools and extend the assessment - to solve the few problems identified on present computing tools and extend the assessment - to perform safety studies on the basis of plant operation feedback - to contribute to treating the safety issues related to the future reactors and in particular the case of severe accidents which have to be taken into account from the design stage. The program on severe accidents is aimed to support the design studies performed by the industrial partners and to provide computing tools which model the various phases of severe accidents and will be validated on experiments performed with real and simulating materials. The main lines of the program are: - the development of the TOLBIAC 3D code for the thermal hydraulics of core melt pools, which will be validated against the Bali experiment presently under construction - the Sultan experiment, to study the capability of cooling by external flooding of the reactor vessel - the development of the MC-3D code for core melt

  18. Characterization of Foam Catalysts as Packing for Tubular Reactors.

    Czech Academy of Sciences Publication Activity Database

    Lali, Farzad

    2016-01-01

    Roč. 105, JUL 2016 (2016), s. 1-9 ISSN 0255-2701 Institutional support: RVO:67985858 Keywords : overall mass transfer * foam catalyst * tubular reactor Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 2.234, year: 2016

  19. 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

  20. Reactor neutrinos study: integration and characterization of the Nucifer detector

    International Nuclear Information System (INIS)

    Gaffiot, Jonathan

    2012-01-01

    The major advances done in the understanding of neutrinos properties and in detector technology have opened the door to a new discipline: the Applied Antineutrino Physics. Indeed, this particle has the great advantage to carry information from its emission place without perturbation. Because neutrinos are inextricably linked to nuclear processes, new applications are in nuclear safeguards. In this context, the Nucifer project aims to test a small electron-antineutrino detector to be installed a few 10 meters from a reactor core for monitoring its thermal power and for testing the sensitivity to the plutonium content. Moreover, recent re-analysis of previous short-distance reactor-neutrino experiments shows a significant discrepancy between measured and expected neutrino count rates. Among the various hypotheses a new phenomenon as the existence of a fourth sterile neutrino can explain this anomaly. To be able to count neutrinos and get the corresponding energy spectrum, the detection is based on the inverse beta decay in about 850 kg of doped liquid scintillator. The experimental challenge is to operate such a small detector in a high background place, due to the closeness with the surface and the reactor radiations. The detector is now finished and data taking has begun at the Osiris research reactor in Saclay since April 2012. Sadly, unexpected low liquid attenuation length and high gamma background level prevented us to highlight neutrinos. We are now waiting for a liquid change and a new lead wall to study reactor monitoring and to test the sterile neutrino hypothesis. (author) [fr

  1. Reactor Materials Program probability of indirectly--induced failure of L and P reactor process water piping

    International Nuclear Information System (INIS)

    Daugherty, W.L.

    1988-01-01

    The design basis accident for the Savannah River Production Reactors is 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 material. The Reactor Materials Program was initiated to provide the technical basis for an alternate credible design basis accident. One aspect of this work is to determine the probability of the DEGB; to show that in addition to being incredible, it is also highly improbable. The probability of a DEGB is broken into two parts: failure by direct means, and indirectly-induced failure. Failure of the piping by direct means can only be postulated to occur if an undetected crack grows to the point of instability, causing a large pipe break. While this accident is not as severe as a DEGB, it provides a conservative upper bound on the probability of a direct DEGB of the piping. The second part of this evaluation calculates the probability of piping failure by indirect causes. Indirect failure of the piping can be triggered by an earthquake which causes other reactor components or the reactor building to fall on the piping or pull it from its supports. Since indirectly-induced failure of the piping will not always produce consequences as severe as a DEGB, this gives a conservative estimate of the probability of an indirectly- induced DEGB. This second part, indirectly-induced pipe failure, is the subject of this report. Failure by seismic loads in the piping itself will be covered in a separate report on failure by direct causes. This report provides a detailed evaluation of L reactor. A walkdown of P reactor and an analysis of the P reactor building provide the basis for extending the L reactor results to P reactor

  2. The plutonium recycle for PWR reactors from brazilian nuclear program

    International Nuclear Information System (INIS)

    Rubini, L.A.

    1978-01-01

    The purpose of this thesis is to evaluate the material requirements of the nuclear fuel cycle with plutonium recycle. The study starts with the calculation of a reference reactor and has flexibility to evaluate the 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. 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 fifth operation cycle of the thermal reactors. (author)

  3. Characterization of filter cartridges from the IEA-R1 reactor by radiochemical method

    International Nuclear Information System (INIS)

    Geraldo, Bianca; Vicente, Roberto; Ferreira, Robson J.; Goes, Marcos M.; Marumo, Julio T.

    2015-01-01

    The filter cartridges used in water purification system of research nuclear reactor IEA-R1 are considered radioactive wastes after their useful life. The characterization of these wastes is one of the stages of management, which aims to identify and quantify the radionuclides present, including those known as 'difficult to measure' (DTM) radionuclides. Establish a radiochemical analysis methodology for this type of waste is a difficult job, not only by the application of these techniques, but also by the amount of radionuclides that should be analyzed. In the waste produced in a nuclear reactor, the most important radionuclides are fission products, activation products and transuranic elements. Since these radionuclides emit gamma radiation not measurable in its decay process and consequently are difficult to measure, their concentrations can be estimated by indirect methods such as scale factors. This method is used to evaluate the DTM concentration, which is represented by alpha and beta nuclides using the correlation between them and the radionuclide key, a gamma emitter. The objective of this work is to describe a radiochemical analysis methodology for gamma emitter nuclides, present in the filter cartridges, evaluating the activity and concentrations by destructive assays. At the same time, two studies have been performed by non-destructive assays, the first one based on dose rates and the point kernel method to correlate the results and the second one based on calibration efficiency with Monte Carlo method. These studies belong to the radioactive waste characterization program that has been conducted at the Waste Management Laboratory of Nuclear and Energy Research Institute, IPEN-CNEN/SP. (author)

  4. Characterization of filter cartridges from the IEA-R1 reactor by radiochemical method

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Bianca; Vicente, Roberto; Ferreira, Robson J.; Goes, Marcos M.; Marumo, Julio T., E-mail: bgeraldo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The filter cartridges used in water purification system of research nuclear reactor IEA-R1 are considered radioactive wastes after their useful life. The characterization of these wastes is one of the stages of management, which aims to identify and quantify the radionuclides present, including those known as 'difficult to measure' (DTM) radionuclides. Establish a radiochemical analysis methodology for this type of waste is a difficult job, not only by the application of these techniques, but also by the amount of radionuclides that should be analyzed. In the waste produced in a nuclear reactor, the most important radionuclides are fission products, activation products and transuranic elements. Since these radionuclides emit gamma radiation not measurable in its decay process and consequently are difficult to measure, their concentrations can be estimated by indirect methods such as scale factors. This method is used to evaluate the DTM concentration, which is represented by alpha and beta nuclides using the correlation between them and the radionuclide key, a gamma emitter. The objective of this work is to describe a radiochemical analysis methodology for gamma emitter nuclides, present in the filter cartridges, evaluating the activity and concentrations by destructive assays. At the same time, two studies have been performed by non-destructive assays, the first one based on dose rates and the point kernel method to correlate the results and the second one based on calibration efficiency with Monte Carlo method. These studies belong to the radioactive waste characterization program that has been conducted at the Waste Management Laboratory of Nuclear and Energy Research Institute, IPEN-CNEN/SP. (author)

  5. Education program at the Massachusetts Institute of Technology research reactor for pre-college science teachers

    International Nuclear Information System (INIS)

    Hopkins, G.R.; Fecych, W.; Harling, O.K.

    1989-01-01

    A Pre-College Science Teacher (PCST) Seminar program has been in place at the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory for 4 yr. The purpose of the PCST program is to educate teachers in nuclear technology and to show teachers, and through them the community, the types of activities performed at research reactors. This paper describes the background, content, and results of the MIT PCST program

  6. 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

  7. PODESY program for flux mapping of CNA II reactor:

    International Nuclear Information System (INIS)

    Ribeiro Guevara, Sergio

    1988-01-01

    The PODESY program, developed by KWU, calculates the spatial flux distribution of CNA II reactor through a three-dimensional expansion of 90 incore detector measurements. The calculation is made in three steps: a) short-term calculation which considers the control rod positions and it has to be done each time the flux mapping is calculated; b) medium-term calculation which includes local burn-up dependent calculation made by diffusion methods in macro-cell configurations (seven channels in hexagonal distribution), and c) long-term calculation, or macroscopic flux determination, that is a fitting and expansion of measured fluxes, previously corrected by local effects, using the eigen functions of the modified diffusion equation. The paper outlines development of step (c) of the calculation. The incore detectors have been located in the central zone of the core. In order to obtain low errors in the expansion procedure it is necessary to include additional points, whose flux values are assumed to be equivalent to detector measurements. These flux values are calculated with detector measurements and a spatial flux distribution calculated by a PUMA code. This PUMA calculation employs a smooth burn-up distribution (local burn-up variations are considered in step (b) of the whole calculation) representing the state of core evolution at the calculation time. The core evolution referred to ends when the equilibrium core condition is reached. Additionally, a calculation method to be employed in the plant in case of incore detector failures, is proposed. (Author) [es

  8. Design reliability assurance program for Korean next generation reactor

    International Nuclear Information System (INIS)

    Lee, Beom-Su; Han, Jin-Kyu; Na, Jang Hwan; Yoo, Kyung Yeong

    1997-01-01

    The Korean Next Generation Reactor (KNGR) project is to develop standardized nuclear power plant design for the construction of future nuclear power plants in Korea. The main purpose of the KNGR project is to develop the advanced nuclear power plants, which enhance safety and economics significantly through the incorporation of design concepts for severe accident prevention and mitigation, supplementary passive safety concept, simplification and application of modularization and so on. For those, Probabilistic Safety Assessment (PSA) and availability study will be performed at the early stage of the design, and the Design Reliability Assurance Program (D-RAP) is applied in the development of the KNGR to ensure that the safety and availability evaluated in the PSA and availability study at the early phase of the design is maintained through the detailed design, construction, procurement and operation of the plants. This paper presents the D-RAP concept that could be applied at the stage of the basic design of the nuclear power plants, based on the models for the reference plants and/or similar plants. 4 refs., 1 fig

  9. 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

  10. 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)

  11. 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.

  12. Characterizing the tribological behaviour of fast breeder reactor materials

    International Nuclear Information System (INIS)

    Depierre, J.; Raffailhac, J.

    1984-04-01

    The object of these tests is to define the behaviour of material couples working in conditions as representative as possible of reactor operation. For this purpose a certain number of test installations have been developed to simulate the most typical cases of friction encountered: plane to plane geometry, rotational bearings, guiding bearings. Endurance tests have also been carried out on ball bearings and ballscrews samples. As said before, the test conditions attempt to reproduce as faithfully as possible the environment of the materials used in fast breeder reactors, particularly in: - using purified liquid sodium, and maintaining it isotherm, respectively at three temperature levels: 180, 400 and 550 0 C; - or using argon containing sodium aerosol particles. Some typical values of friction coefficients and rates of wear obtained during the tests with certain couples of materials are given here as examples. The aims which are currently guiding the direction of the tests are also briefly described

  13. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Alvarez, I.; Herrera, E.; Lima, L.; Tores, J.; Lopez, M.C.; Ixquiac, M.

    1996-01-01

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K o neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott's formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented

  14. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Rizo, O; Alvarez, I; Herrera, E; Lima, L; Tores, J [Secretaria Ejecutiva para Asuntos Nucleares, Holguin (Cuba). Delegacion Territorial; Manso, M V [Centro de Isotopos, La Habana (Cuba); Lopez, M C [Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico); Ixquiac, M [Universidad de San Carlos de Guatemala, Guatemala City (Guatemala)

    1997-12-31

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K{sub o} neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott`s formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented.

  15. Development of a computer program for the simulation of one-dimensional fixed- and moving-bed reactors

    International Nuclear Information System (INIS)

    Hartner, P.

    1996-11-01

    Chemical reactors with a flow through a bed of solid particles are of great importance in the processing industry. Modern computational tools allow for an improved characterization of the complex facts in such reactors leading to new opportunities of optimizing the reactor operation and environmental effects. This thesis is concerned with the development of the one-dimensional simulation software REASIM. The program covers the effects within a reacting bed and is designed for fixed and moving beds. To describe the reactor the balances for energy, momentum and mass are solved. The drying of the particles, pyrolysis and chemical gas-solid and gas-gas reactions are considered. For the description of the chemical gas-solid reactions a particle model for porous solids is developed. The calculation of mass transfer and of chemical reactions is strictly separated. All parameters necessary for the model can be measured in the laboratory. The model equations form a system of partial differential equations. This system is transformed to a set of ordinary differential equations. It is found that the best discretization method is the method of finite differences with the upwind-scheme for situations where convection is strong. The program has a modular structure making it is easy to replace parts of the program by new, improved modules if they become available. (author)

  16. Radiological control aspects of the fabrication of the Light Water Breeder Reactor core (LWBR Development Program)

    International Nuclear Information System (INIS)

    Schultz, B.G.

    1979-05-01

    A description is presented of the radiological control aspects of the fabrication of the Light Water Breeder Reactor (LWBR) core. Included are the radiological control criteria applied for the design and use of fabrication facilities, the controls and limits imposed to minimize radiaion exposure to personnel, and an evaluation of the applied radiological program in meeting the program objectives. The goal of the LWBR program is to develop the technology to breed in light water reactors so that nuclear fuel may be used significantly more efficiently in these reactors. This technology is being developed by designing and fabricating a breeder reactor core, utilizing thoria (ThO 2 ) and binary thoria--urania (ThO 2 - 233 UO 2 ) fuel, to be operated in the existing pressurized water reactor plant owned by the Department of Energy at Shippingport, Pennsylvania

  17. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

    1980-03-01

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  18. University Reactor Sharing Program. Final report, September 30, 1992--September 29, 1994

    International Nuclear Information System (INIS)

    Wehring, B.W.

    1995-01-01

    Over the past 20 years, the number of nuclear reactors on university campuses in the US declined from more than 70 to less than 40. Contrary to this trend, The University of Texas at Austin constructed a new reactor facility at a cost of $5.8 million. The new reactor facility houses a new TRIGA Mark II reactor which replaces an in-ground TRIGA Mark I reactor located in a 50-year old building. The new reactor facility was constructed to strengthen the instruction and research opportunities in nuclear science and engineering for both undergraduate and graduate students at The University of Texas. On January 17, 1992, The University of Texas at Austin received a license for operation of the new reactor. Initial criticality was achieved on March 12, 1992, and full power operation, on March 25, 1992. The UT-TRIGA research reactor provides hands-on education, multidisciplinary research and unique service activities for academic, medical, industrial, and government groups. Support by the University Reactor Sharing Programs increases the availability of The University of Texas reactor facility for use by other educational institutions which do not have nuclear reactors

  19. Double-digit growth characterizes 1981 nuclear programs abroad

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    A 1980 international reactor survey shows a doubling of orders for nuclear power programs outside the US and a nearly 14% increase in nuclear capacity abroad. US manufacturers will not fill any of the 10 new foreign orders of 1980, and they received no new domestic orders. A statistical breakdown lists plants by country, type, net megawatts, year of reactor order, reactor manufacturer, location, date of commercial operation, and primary utility. The report also summarizes fuel-reprocessing and waste-management development in key countries. 1 figure, 1 table

  20. CFARMHD -- A MathCAD PC program to evaluate performance and economics of CFARII fusion reactors

    International Nuclear Information System (INIS)

    Logan, B.G.

    1992-01-01

    This report describes a PC computer program ''CFARMHD'', developed to evaluate the performance (MHD cycle efficiency) and economics (Cost-of-Electricity CoE) for pulsed fusion reactors using the Compact Fusion Advanced Rankine II (CFARII) MHD Balance of Plant (BoP). The CFARII concept to which this code applies is generic to any fusion driver which can be characterized by an assumed yield Y (GJ), target gain G, and unit cost ($/joule driver). The CFARMHD code models the sizes, masses, energies, mass flows and powers corresponding to the physical systems and optimizes them to minimize CoE for given Y, G, $/joule, and choice of material for the working fluid (cast as solid spherical shells around the target). A description of the models used in the CFARMHD code is given in Section 11, and the CoE minimization procedure used in the code is described in Section III

  1. Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Fromm, Bradley [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hauch, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sridharan, Kumar [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-01

    A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.

  2. Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

    International Nuclear Information System (INIS)

    Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

    2016-01-01

    A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulation tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.

  3. Advanced Characterization Techniques for Silicon Carbide and Pyrocarbon Coatings on Fuel Particles for High Temperature Reactors (HTR)

    Energy Technology Data Exchange (ETDEWEB)

    Basini, V.; Charollais, F. [CEA Cadarache, DEN/DEC/SPUA, BP 1, 13108 St Paul Lez Durance (France); Dugne, O. [CEA Marcoule, DEN/DTEC/SCGS BP 17171 30207 Bagnols sur Ceze (France); Garcia, C. [Laboratoire des Composites Thermostructuraux (LCTS), UMR CNRS 5801, 3 allee de La Boetie, 33600 Pessac (France); Perez, M. [CEA Grenoble DRT/DTH/LTH, 17 rue des Martyrs, 38054 Grenoble cedex 9 (France)

    2008-07-01

    Cea and AREVA NP have engaged an extensive research and development program on HTR (high temperature reactor) fuel. The improving of safety of (very) high temperature reactors (V/HTR) is based on the quality of the fuel particles. This requires a good knowledge of the properties of the four-layers TRISO particles designed to retain the uranium and fission products during irradiation or accident conditions. The aim of this work is to characterize exhaustively the structure and the thermomechanical properties of each unirradiated layer (silicon carbide and pyrocarbon coatings) by electron microscopy (SEM, TEM), selected area electronic diffraction (SEAD), thermo reflectance microscopy and nano-indentation. The long term objective of this study is to define pertinent parameters for fuel performance codes used to better understand the thermomechanical behaviour of the coated particles. (authors)

  4. Characterization of 14C in Swedish light water reactors.

    Science.gov (United States)

    Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

    2008-08-01

    This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

  5. Characterization of the TRIGA Mark II reactor full-power steady state

    Energy Technology Data Exchange (ETDEWEB)

    Cammi, Antonio, E-mail: antonio.cammi@polimi.it [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Zanetti, Matteo [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica [University of Milano-Bicocca, Physics Department “G. Occhialini” and INFN Section, Piazza dell’Ateneo Nuovo, 20126 Milan (Italy); Magrotti, Giovanni; Prata, Michele; Salvini, Andrea [University of Pavia, Applied Nuclear Energy Laboratory (L.E.N.A.), Via Gaspare Aselli 41, 27100 Pavia (Italy)

    2016-04-15

    Highlights: • Full-power steady state characterization of the TRIGA Mark II reactor. • Monte Carlo and Multiphysics simulation of the TRIGA Mark II reactor. • Sub-cooled boiling effects in the TRIGA Mark II reactor. • Thermal feedback effects in the TRIGA Mark II reactor. • Experimental data based validation. - Abstract: In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the “Multiphysics” model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics.

  6. Water reactor fuel characterization. Part of a coordinated programme on non-destructive techniques for reactor fuel characterization

    International Nuclear Information System (INIS)

    Levai, F.

    1983-06-01

    The report describes an optical/mechanical system of examining nuclear reactor fuel bundles by tomographic imaging using high contrast X-ray film. This low cost system does not use expensive detectors or digital computers. The apparatus assembled from ordinary and available components consists of a 2π scanner, a back projector and filters. Although the system described and tested is based on transmission tomography, the report also discusses the extension of the concept to emission tomography

  7. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1982-01-01

    The fast breeder reactor development project in Japan has been in progress for the past twelve months and will be continued this fiscal year, from April 1982 through March 1983, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1981. The 1982 year budget for R and D work and for construction of a prototype fast breeder reactor MONJU is approximately 20 and 27 billion yen respectively, excluding wages for the personnel of the Power Reactor and Nuclear Fuel Development Corporation, PNC. The number of the technical people currently engaged in the fast breeder reactor development in the PNC is approximately 530, excluding those working for plutonium fuel fabrication. Concerning the experimental fast reactor JOYO, power increase from 50 MWt to 75 MWt was made in July 1979 and six operational cycles at 75 MWt were completed in December 1981. With respect to the prototype reactor MONJU, progress toward construction has been made and an environmental impact statement of the reactor was approved by the authorities concerned, and the licensing of the first step was completed at the end of 1981. Preliminary design studies of a large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  8. Health and Safety Plan for Operations Performed for the Environmental Restoration Program: Task, Characterization of Potential Waste Sources at Auxiliary Reactor Area-1 Operable Unit 5--07 site ARA-02

    International Nuclear Information System (INIS)

    Pickett, S.L.; Morton, S.L.

    1992-06-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG ampersand G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the ERP. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP

  9. Health and Safety Plan for Operations Performed for the Environmental Restoration Program: Task, Characterization of Potential Waste Sources at Auxiliary Reactor Area-1 Operable Unit 5--07 site ARA-02

    Energy Technology Data Exchange (ETDEWEB)

    Pickett, S.L.; Morton, S.L.

    1992-06-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the ERP. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP.

  10. Health and Safety Plan for Operations Performed for the Environmental Restoration Program: Task, Characterization of Potential Waste Sources at Auxiliary Reactor Area-1 Operable Unit 5--07 site ARA-02

    Energy Technology Data Exchange (ETDEWEB)

    Pickett, S.L.; Morton, S.L.

    1992-06-01

    This document constitutes the generic health and safety plan for the Environmental Restoration Program (ERP). It addresses the health and safety requirements of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA); Occupational Safety and Health Administration (OSHA) 29 CFR 1910.120 standard; and EG&G Idaho, Inc. This plan is a guide to individuals who must complete a health and safety plan for a task performed for the ERP. It contains a task specific addendum that, when completed, specifically addresses task specific health and safety issues. This health and safety plan reduces the time it takes to write a task specific health and safety plan by providing discussions of requirements, guidance on where specific information is located, and specific topics in the Addendum that must be discussed at a task level. This format encourages a complete task specific health and safety plan and a standard for all health and safety plans written for ERP.

  11. Developing a framework for a sustainable research reactor program in the UAE

    Energy Technology Data Exchange (ETDEWEB)

    Almarri, Khalid [Senate Member and Representative of Students of PhD Program in Project Management, Ajman (United Arab Emirates)

    2013-07-01

    In 2009, the UAE passed a milestone of preparations for the involvement in the nuclear area by awarding its first nuclear power plant. And to maximise its nuclear benefits, the country needs to develop a research reactor program. This paper's objectives are: a) Selecting and analysing model case studies from other nations to establish the success factors of research reactors in the UAE, and the preparedness of local industries to maximise the potential of the reactors. b) Establishing how the UAE's research reactors will contribute to the coalition and sharing of good practices with other countries as promoted by IAEA.

  12. 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

  13. 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.

  14. 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.

  15. Nuclear power reactor analysis, methods, algorithms and computer programs

    International Nuclear Information System (INIS)

    Matausek, M.V

    1981-01-01

    Full text: For a developing country buying its first nuclear power plants from a foreign supplier, disregarding the type and scope of the contract, there is a certain number of activities which have to be performed by local stuff and domestic organizations. This particularly applies to the choice of the nuclear fuel cycle strategy and the choice of the type and size of the reactors, to bid parameters specification, bid evaluation and final safety analysis report evaluation, as well as to in-core fuel management activities. In the Nuclear Engineering Department of the Boris Kidric Institute of Nuclear Sciences (NET IBK) the continual work is going on, related to the following topics: cross section and resonance integral calculations, spectrum calculations, generation of group constants, lattice and cell problems, criticality and global power distribution search, fuel burnup analysis, in-core fuel management procedures, cost analysis and power plant economics, safety and accident analysis, shielding problems and environmental impact studies, etc. The present paper gives the details of the methods developed and the results achieved, with the particular emphasis on the NET IBK computer program package for the needs of planning, construction and operation of nuclear power plants. The main problems encountered so far were related to small working team, lack of large and powerful computers, absence of reliable basic nuclear data and shortage of experimental and empirical results for testing theoretical models. Some of these difficulties have been overcome thanks to bilateral and multilateral cooperation with developed countries, mostly through IAEA. It is the authors opinion, however, that mutual cooperation of developing countries, having similar problems and similar goals, could lead to significant results. Some activities of this kind are suggested and discussed. (author)

  16. Artificial intelligence program in a computer application supporting reactor operations

    International Nuclear Information System (INIS)

    Stratton, R.C.; Town, G.G.

    1985-01-01

    Improving nuclear reactor power plant operability is an ever-present concern for the nuclear industry. The definition of plant operability involves a complex interaction of the ideas of reliability, safety, and efficiency. This paper presents observations concerning the issues involved and the benefits derived from the implementation of a computer application which combines traditional computer applications with artificial intelligence (AI) methodologies. A system, the Component Configuration Control System (CCCS), is being installed to support nuclear reactor operations at the Experimental Breeder Reactor II

  17. 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

  18. LTFR-4, Library Generated for Fast Reactor Design Program from JAERI Fast-Set Multigroup Constant

    International Nuclear Information System (INIS)

    Suzuki, Tomoo

    1971-01-01

    Nature of physical problem solved: The program processes JAERI-Fast group constants sets of less than 30-group and prepares a binary library tape for efficient usage by a series of related fast reactor design calculation programmes

  19. Reactor vessel assessment and the development of a reactor vessel life extension program for Calvert Cliffs Units One and Two

    International Nuclear Information System (INIS)

    Montgomery, B.; Hijeck, P.J.

    1988-01-01

    A study has been undertaken to provide a general assessment of the life extension capabilities for the Calvert Cliffs Units One and Two reactor pressure vessels. The purpose of the study is to assess the general life extension capabilities for the Calvert Cliffs reactor pressure vessels based upon an extension and variation of the Surry pilot plant life extension study. This assessment provided a detailed reactor vessel surveillance program for plant life extension along with a hierarchy of specific tasks necessary for attaining maximum useful life. The assessment identified a number of critical issues which may impact life attainment and extension along with potential solutions to address these issues to ensure the life extension option is not precluded

  20. 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...

  1. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  2. Overview of the fast reactors fuels program. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

    1980-04-01

    Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

  3. 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

  4. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [Unidad de Actividad Combustibles Nucleares Comision Nacional de Energia Atomica (CNE4), Avda. del Libertador, 8250 C1429BNO Buenos Aires (Argentina)

    2002-07-01

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm{sup 3}. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  5. Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

    International Nuclear Information System (INIS)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H.

    2002-01-01

    One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm 3 . PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

  6. 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)

  7. Sodium components cleaning status in the Italian fast reactor program

    Energy Technology Data Exchange (ETDEWEB)

    De Luca, B [CNEN-RIT/MAT - Laboratorio Sviluppo Processi - C.S.N. Cassacia, Rome (Italy); Labanti, V [CNEN-DRV, Bologna (Italy); Mennucci, M [NIRA, Genoa (Italy)

    1978-08-01

    As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed.

  8. Summary of particle bed reactor designs for the Space Nuclear Thermal Propulsion Program

    Science.gov (United States)

    Powell, J. R.; Ludewig, H.; Todosow, M.

    1993-09-01

    A summary report of the Particle Bed Reactor (PBR) designs considered for the space nuclear thermal propulsion program has been prepared. The first chapters outline the methods of analysis, and their validation. Monte Carlo methods are used for the physics analysis, several new algorithms are used for the fluid dynamics heat transfer and engine system analysis, and commercially available codes are used for the stress analysis. A critical experiment, prototypic of the PBR was used for the physics validation, and blowdown experiments using fuel beds of prototypic dimensions were used to validate the power extraction capabilities from particle beds. In all four different PBR rocket reactor designs were studied to varying degrees of detail. They varied in power from 400 MW to 2000 MW. These designs were all characterized by a negative prompt coefficient, due to Doppler feedback, and the feedback due to moderator heat up varied from slightly negative to slightly positive. In all practical cases, the coolant worth was positive, although core configurations with negative coolant worth could be designed. In all practical cases the thrust/weight ratio was greater than 20.

  9. Data-processing program from the operating modes of the nuclear reactor (P0DER)

    International Nuclear Information System (INIS)

    Totev, T.L.; Boyadzhiev, A.I.

    1981-01-01

    A program PODER for processing data from the operating modes of the reactors taking into account the effects of corrosion, hydration, and deformation of the nuclear reactor fuel element sheathing, the formation of the corrosion product deposits, the change in the geometric dimensions of the nuclear reactor fuel element due to the temperature deformation, as well as the various gas fillers, are elaborated. The ''hot channel'' method determining the reliability of the system is realized. The basic equations describing the thermohydraulic processes in nuclear reactors are solved by the finite difference method. Approximations are carried out with the approach of least squares. The temperature distribution versus the zirconium sheathing height is computed for the case of WWER-440 type reactors. The advantages of the proposed program P0DER are discussed

  10. 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)

  11. 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)

  12. Integration of improved decontamination and characterization technologies in the decommissioning of the CP-5 research reactor

    International Nuclear Information System (INIS)

    Bhattacharyya, S. K.; Boing, L. E.

    2000-01-01

    The aging of research reactors worldwide has resulted in a heightened awareness in the international technical decommissioning community of the timeliness to review and address the needs of these research institutes in planning for and eventually performing the decommissioning of these facilities. By using the reactors already undergoing decommissioning as test beds for evaluating enhanced or new/innovative technologies for decommissioning, it is possible that new techniques could be made available for those future research reactor decommissioning projects. Potentially, the new technologies will result in: reduced radiation doses to the work force, larger safety margins in performing decommissioning and cost and schedule savings to the research institutes in performing the decommissioning of these facilities. Testing of these enhanced technologies for decontamination, dismantling, characterization, remote operations and worker protection are critical to furthering advancements in the technical specialty of decommissioning. Furthermore, regulatory acceptance and routine utilization for future research reactor decommissioning will be assured by testing and developing these technologies in realistically contaminated environments prior to use in the research reactors. The decommissioning of the CP-5 Research Reactor is currently in the final phase of dismantlement. In this paper the authors present results of work performed at Argonne National Laboratory (ANL) in the development, testing and deployment of innovative and/or enhanced technologies for the decommissioning of research reactors

  13. 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

  14. 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...

  15. N Reactor thermal plume characterization during Pu-only mode of operation

    Energy Technology Data Exchange (ETDEWEB)

    Ecker, R.M.; Thompson, F.L.; Whelan, G.

    1983-04-01

    Pacific Northwest Laboratories (PNL) performed field and modeling studies -from March 1982 through June 1983 to characterize the thermal plume from the N Reactor heated water outfall while the N Reactor operated in the Pu-only mode. Part 1 of this report deals with the field studies conducted to characterize the N Reactor thermal plume while in the Pu-only mode of operation. It includes a description of the study area, a description of field tasks and procedures, and data collection results and discussion. Part 2 describes the computer simulation of the thermal plume under different flow conditions and the calibration of the model used. It includes a description of the computer model and the assumptions on which it is based, a presentation of the input data used in this application, and a discussion of modeling results. Because the field studies were restricted by the NPOES permit variance to the spring months when high Columbia River flows prevail the mathematical modeling of the N Reactor thermal plume while the reactor operates in the Pu-only mode is instrumental in characterizing the plume during low Columbia River flows.

  16. A transparent Pyrex μ-reactor for combined in situ optical characterization and photocatalytic reactivity measurements

    International Nuclear Information System (INIS)

    Dionigi, F.; Hansen, O.; Nielsen, M. G.; Chorkendorff, I.; Vesborg, P. C. K.; Pedersen, T.

    2013-01-01

    A new Pyrex-based μ-reactor for photocatalytic and optical characterization experiments is presented. The reactor chamber and gas channels are microfabricated in a thin poly-silicon coated Pyrex chip that is sealed with a Pyrex lid by anodic bonding. The device is transparent to light in the UV-vis-near infrared range of wavelengths (photon energies between ∼0.4 and ∼4.1 eV). The absorbance of a photocatalytic film obtained with a light transmission measurement during a photocatalytic reaction is presented as a proof of concept of a photocatalytic reactivity measurement combined with in situ optical characterization. Diffuse reflectance measurements of highly scattering photocatalytic nanopowders in a sealed Pyrex μ-reactor are also possible using an integrating sphere as shown in this work. These experiments prove that a photocatalyst can be characterized with optical techniques after a photocatalytic reaction without removing the material from the reactor. The catalyst deposited in the cylindrical reactor chamber can be illuminated from both top and bottom sides and an example of application of top and bottom illumination is presented

  17. Characterization of noise sources in nuclear power reactors

    International Nuclear Information System (INIS)

    Andhill, Gustav

    2004-03-01

    Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D

  18. Characterization of noise sources in nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Andhill, Gustav

    2004-03-01

    Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D.

  19. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  20. Characterization of Zircaloy-4 tubing procured for fuel cladding research programs

    International Nuclear Information System (INIS)

    Chapman, R.H.

    1976-01-01

    A quantity of Zircaloy-4 tubing [10.92 mm outside diameter by 0.635 mm wall thickness] was purchased specifically for use in a number of related fuel cladding research programs sponsored by the Division of Reactor Safety Research, Nuclear Regulatory Commission (NRC/RSR). Identical tubing (produced simultaneously and from the same ingot) was purchased concurrently by the Electric Power Research Institute (EPRI) for use in similar research programs sponsored by that organization. In this way, source variability and prior fabrication history were eliminated as parameters, thus permitting direct comparison (as far as as-received material properties are concerned) of experimental results from the different programs. The tubing is representative of current reactor technology. Consecutive serial numbers assigned to each tube identify the sequence of the individual tubes through the final tube wall reduction operation. The report presented documents the procurement activities, provides a convenient reference source of manufacturer's data and tubing distribution to the various users, and presents some preliminary characterization data. The latter have been obtained routinely in various research programs and are not complete. Although the number of analyses, tests, and/or examinations performed to date are insufficient to draw statistically valid conclusions with regard to material characterization, the data are expected to be representative of the as-received tubing. It is anticipated that additional characterizations will be performed and reported routinely by the various research programs that use the tubing

  1. FFTF scale-model characterization of flow-induced vibrational response of reactor internals

    International Nuclear Information System (INIS)

    Ryan, J.A.; Julyk, L.J.

    1977-01-01

    As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36 percent to 111 percent of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable

  2. FFTF scale-model characterization of flow induced vibrational response of reactor internals

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, J A; Julyk, L J [Hanford Engineering Development Laboratory, Richland, WA (United States)

    1977-12-01

    As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

  3. FFTF scale-model characterization of flow induced vibrational response of reactor internals

    International Nuclear Information System (INIS)

    Ryan, J.A.; Julyk, L.J.

    1977-01-01

    As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

  4. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  5. A review of fast reactor program in Japan - April 1984

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1984-01-01

    The fast breeder reactor development project in PNC has been in progress steadily in these eighteen years. Concerning the experimental fast reactor, JOYO, the MK-II core attained criticality on November 22, 1982 with 51 fuel assemblies, and received the ''Certificate of Inspection before Operation'' from Government Authority on March 31, 1983, after 100 hours operation with the rated output of 100 MW. Since then, the core has been utilized to implement irradiation bed characteristics test, and to irradiate fuels and structural materials especially for the prototype reactor MONJU. With respect to the prototype reactor MONJU, the installation permit was issued on May 27, 1983, from the prime minister, and the contracts of the first stage between PNC and fabricators were made recently. At the same time, almost all the licenses of preparatory construction works were issued by March 1983, and preparatory construction works were started in April 1983. On the other hand, conceptual design of a demonstration reactor is now under way in a close cooperation with concerned authorities and utilities, as well as investigations of the way of conducting necessary research and development

  6. A database model for the radiological characterization of the RA reactor in the 'Vinca' Institute

    International Nuclear Information System (INIS)

    Steljic, M.; Ljubenov, V.

    2004-01-01

    During the preparation and realization of the radiological characterization of nuclear facility it is necessary to organize, store, review and process large amount of various data types. The documentation has to be treated according to the quality assurance (QA) programme requirements, and to ultimate goal would be to establish the unique record management system (RMS) for the nuclear facility decommissioning project. This paper presents the design details of the database model for the radiological characterization of the RA research reactor (author) [sr

  7. Plutonium recycle test reactor characterization activities and results

    International Nuclear Information System (INIS)

    Cornwell, B.C.

    1997-01-01

    Report contains results of PRTR core and associated structures characterization performed in January and February of 1997. Radiation survey data are presented, along with recommendations for stabilization activities before transitioning to a decontamination and decommissioning function. Recommendations are also made about handling the waste generated by the stabilization activities, and actions suggested by the Decontamination and Decommissioning organization

  8. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion

    International Nuclear Information System (INIS)

    Marques, J.G.; Sousa, M.; Santos, J.P.; Fernandes, A.C.

    2011-01-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1 MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  9. Characterization of biofilm in 200W fluidized bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Michelle H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Saurey, Sabrina D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Lee, Brady D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Parker, Kent E. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Eisenhauer, Emalee E. R. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Cordova, Elsa A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Golovich, Elizabeth C. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

    2014-09-29

    Contaminated groundwater beneath the 200 West Area at the Hanford Site in Southeast Washington is currently being treated using a pump and treat system to remove organics, inorganics, radionuclides, and metals. A granular activated carbon-based fluidized bed reactor (FBR) has been added to remove nitrate, hexavalent chromium and carbon tetrachloride. Initial analytical results indicated the microorganisms effectively reduced many of the contaminants to less than cleanup levels. However shortly thereafter operational upsets of the FBR include carbon carry over, over production of microbial extracellular polymeric substance (biofilm) materials, and over production of hydrogen sulfide. As a result detailed investigations were undertaken to understand the functional diversity and activity of the microbial community present in the FBR over time. Molecular analyses including terminal restriction fragment length polymorphism analysis, quantitative polymerase chain reaction and fluorescent in situ hybridization analyses were performed on the microbial community extracted from the biofilm within the bed and from the inoculum, to determine functional dynamics of the FBR bed over time and following operational changes. Findings from these analyses indicated: 1) the microbial community within the bed was completely different than community used for inoculation, and was likely from the groundwater; 2) analyses early in the testing showed an FBR community dominated by a few Curvibacter and Flavobacterium species; 3) the final sample taken indicated that the microbial community in the FBR bed had become more diverse; and 4) qPCR analyses indicated that bacteria involved in nitrogen cycling, including denitrifiers and anaerobic ammonia oxidizing bacteria, were dominant in the bed. These results indicate that molecular tools can be powerful for determining functional diversity within FBR type reactors. Coupled with micronutrient, influent and effluent chemistry

  10. 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

  11. 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

  12. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  13. Pilot program: NRC severe reactor accident incident response training manual: Severe reactor accident overview

    International Nuclear Information System (INIS)

    McKenna, T.J.; Martin, J.A.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

    1987-02-01

    This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Severe Reactor Accident Overview is the second in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident assesment. Each volume serves, respectively, as the text for a course of instruction in a series of courses. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

  14. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    International Nuclear Information System (INIS)

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates

  15. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle description

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    The Nonproliferation Alterntive Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates.

  16. A review of fast reactor program in Japan

    International Nuclear Information System (INIS)

    1996-01-01

    The main R and D results of Japanese activities are summarized as follows: (1) the experimental 140 MW(th) sodium cooled fast reactor 'Joyo' provided abundant experimental data and excellent operational records, attaining more than 50,000 hours of operation since its first criticality in 1977; (2) the prototype 280 MW(e) fast reactor 'Monju' reached initial criticality on 5 April 1994; presently Monju is under the cold shutdown state because of secondary sodium leak on 8 December 1995, and multiple cause investigations of the sodium leak are being performed; (3) the Japan Atomic Power Company is promoting design studies for demonstration fast reactor (DFBR) with a power output of 600 MW(e) and R and D for DFBR are being conducted under the cooperation of governmental and private sectors. (author)

  17. A review of fast reactor program in Japan - April 1983

    International Nuclear Information System (INIS)

    Matsuno, Y.

    1983-01-01

    The fast breeder reactor development project in Japan has been in progress during the past twelve months and will be continued in the next fiscal year, from April 1983 through March 1984, at a similar scale of effort both in budget and personnel to those of the fiscal year of 1982. Concerning the experimental fast reactor, JOYO, all the scheduled testings and operations were completed by the end of 1981 and from the beginning of 1982 the change-out work from Mark-I core to Mark-II core has been continued for 11 months. The initial criticality on the Mark-II core was achieved on 22 Nov. 1982 and after 3 months low power physics tests the reactor power was raised up to 100% (100 MWt) in the middle of March 1983. With respect to the prototype reactor MONJU, progress toward construction has been made and the licensing of the second step will be completed in the first half of 1983. Preliminary design studies of large LMFBR are being made by PNC and also by utilities. A design study being conducted by PNC is on a 1000 MWe plant of loop type by extrapolating the technology to be developed by the time of commissioning of MONJU. A group of utilities is conducting a similar study, but covering somewhat wider range of parameters and options of design. Close contact between the group and PNC has been kept. In the future, those design efforts will be combined as a single design effort, when a major effort for developing a large demonstration reactor will be initiated at around the commencement of construction of the prototype reactor MONJU

  18. Characterization of nuclear reactor containment penetrations. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Shackelford, M.H.; Bump, T.R.; Seidensticker, R.W.

    1985-02-01

    This report concludes a preliminary report prepared by ANL for Sandia, published as NUREG/CR-3855, in June 1984. The preliminary report, NUREG/CR-3855, presented the results of a survey of nuclear reactor containment penetrations, covering the number of plants surveyed at that time (22 total). Since that time, an additional 26 plants have been included in the survey. This final report serves two purposes: (1) to add the summary data sheets and penetration details for the additional plants now included in the survey; and (2) to confirm, revise, or add to analyses and discussions presented in the first report which, of course, were based solely on the earlier sample of 22 plants. This final report follows the outline and format of the preliminary survey report. In general, changes and additions to the preliminary report are implied, rather than stated as such to avoid repeated reference to that report. If no changes have been made in a section the title of the section of the previous report is simply repeated followed by ''No Changes''. Some repetition is used for continuity and clarity.

  19. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

  1. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    International Nuclear Information System (INIS)

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions

  2. French studies and research program in pressurized water reactor safety

    International Nuclear Information System (INIS)

    Duco, J.

    1986-06-01

    The aim of researches developed now in France on water reactor safety is to obtain means and knowledge allowing to control accidental situations, including severe situations beyond design basis accidents. The main studies and researches concerning water reactors and described in this report are the following ones: core cooling accident and prevention of severe accidents, fuel behavior in accidental situation, behavior of the containment building, fission product transfer and releases in case of accident, problems related to equipment aging, and, methodology of risk analysis and ''human factor'' studies. Most of these studies follow an analytic approach of phenomena [fr

  3. The industry/EPRI advanced light water reactor program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Sugnet, W.R.; Bilan, W.J.

    1986-01-01

    For the United States nuclear power industry to remain viable, it must be prepared to meet the expected need for new generating capacity in the late 1990s with an improved reactor system. The best hope of meeting this requirement is with evolutionary changes in current LWR systems through system simplification and reevaluation of safety and operational design margins. The grid characteristics and the difficulty in raising capital for large projects indicate that smaller light water reactors (400 to 600 MWe) may play an important role the next generation

  4. Overview of U.S. Fast Reactor Technology Program

    International Nuclear Information System (INIS)

    Hill, Robert

    2013-01-01

    • Concept development studies guide R&D tasks by evaluating system impact for broad variety of technology options: – Small-scale facilities for R&D on key technology; – No near-term plan for demonstration reactor. • Fast reactor R&D is focused on key technologies innovations for performance improvement (cost reduction): – Advanced Structural Materials; – Advanced Energy Conversion; – Advanced Modeling and Simulation. • Other R&D is conducted to address known technology challenges: – Safety and Licensing; – Fuels Development; – Undersodium Viewing

  5. Investigations of the natural fission reactor program. Progress report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Cowan, G.A.; Norris, A.E.

    1978-10-01

    The U.S. study of the Oklo natural reactor began in 1973 with the principal objectives of understanding the processes that produced the reactor and that led to the retention of many of its products. Major facets of the program have been the chemical separation and mass spectrometric analysis of the reactor components and products, the petrological and mineralogical examination of samples taken from the reactor zones, and an interdisciplinary modeling of possible processes consistent with reactor physics, geophysics, and geochemistry. Most of the past work has been on samples taken within the reactor zones. Presently, these studies give greater emphasis to the measurement of mobile products in additional suites of samples collected peripherally and ''downstream'' from the reactor zones. This report summarizes the current status of research and the views of U.S. investigators, with particular reference to the extensive work of the French scientists, concerning the main features of the Oklo natural fission reactor. Also mentioned briefly is the U.S. search for natural fission reactors at other locations

  6. Cross-disciplinary research programs at the Cornell TRIGA reactor

    International Nuclear Information System (INIS)

    Clark, D.D.

    1995-01-01

    This paper describes cross-disciplinary research efforts at the Cornell TRIGA reactor. A new graduate laboratory course for nonspecialists was developed which brought in graduate students from many fields, and a weekly or bimonthly nuclear methods seminars are being held to describe research methods, sample preparation, irradiation, etc

  7. Application of microprocessor based controller in the Breeder Reactor Program

    International Nuclear Information System (INIS)

    Messick, N.C.; Lukas, M.P.

    1985-01-01

    This paper treats Argonne National Laboratory's experience using microprocessor based controllers presently in use on several control loops within the EBR-II reactor facility as well as tests being performed by these controllers. Also included is a discussion of the expandability, modularity, range of capabilities and higher level functions possible using such equipment

  8. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  9. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    1994-01-01

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

  10. NEPTUNIX, a general program of simulation applied to nuclear reactors

    International Nuclear Information System (INIS)

    Bonnemay, A.; Dansac Bon, V.

    1978-01-01

    Most simulation languages admit an incremental description and involve explicit integration algorithms. NEPTUNIX is a simulation language directly admitting algebraic differential equations under an implicit form, and it involves a very efficient implicit integration method with variable step and order. NEPTUNIX is a tool used for building large systems models in the field of nuclear reactors [fr

  11. Programmed elimination of neutronic poisons in nuclear reactors

    International Nuclear Information System (INIS)

    Perriere, G. de la

    1967-11-01

    This work deals with the use of salts of elements having a large neutron capture cross-section, so-called 'soluble poisons' which are dissolved in the moderating water to control the reactivity of heavy-water reactors, and more particularly to compensate the xenon effect in the reactor EL 4. The report describes the controlled elimination of these poisons by fixation on ion-exchange resins. The poisons considered are lithium-6, cadmium and gadolinium in the sulphate form, and boron as boric acid. The thermodynamic and kinetic constants of the ion-exchange reactions were first determined and a study was then made of the fixation of these compounds in beds of small-calibre resins placed in columns. Lithium-6 is the poison which is most easily applicable to compensate the xenon effect in the reactor EL 4. It can be eliminated rapidly and completely from heavy water, and its use does not lead to supplementary problems of protection against the gamma radiation of the reactor circuits. (author) [fr

  12. 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.)

  13. 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)

  14. A dedicated program for the extended longevity of research and training reactors

    International Nuclear Information System (INIS)

    Carriveau, G.W.

    1992-01-01

    In the past 49 years, over 555 research and training reactors have been in operation, with approximately 325 currently in service. The age distribution of operating research reactors shows that the average age is about 24 years; about 74% are 20 years old or older and about 33% are 30 years old or older. This group of reactors represents a very large investment in capital expense with replacement costs in 1990's prices much higher than when they were originally constructed. Furthermore, decommissioning costs may be much greater than the original investments. General Atomics has been directly involved for the better part of the nearly fifty year history of research and training reactors. This paper will describe a General Atomics program illustrating a dedicated commitment to the full service support of extended and improved use for all types of research and training reactors. (author)

  15. 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

  16. Characterization of the molten salt reactor experiment fuel and flush salts

    International Nuclear Information System (INIS)

    Williams, D.F.; Peretz, F.J.

    1996-01-01

    Wise decisions about the handling and disposition of spent fuel from the Molten Salt Reactor Experiment (MSRE) must be based upon an understanding of the physical, chemical, and radiological properties of the frozen fuel and flush salts. These open-quotes staticclose quotes properties can be inferred from the extensive documentation of process history maintained during reactor operation and the knowledge gained in laboratory development studies. Just as important as the description of the salt itself is an understanding of the dynamic processes which continue to transform the salt composition and govern its present and potential physicochemical behavior. A complete characterization must include a phenomenological characterization in addition to the typical summary of properties. This paper reports on the current state of characterization of the fuel and flush salts needed to support waste management decisions

  17. Characterization of the TRIGA Mark III reactor for k0-neutron activation analysis

    International Nuclear Information System (INIS)

    Diaz R, O.; Herrera P, E.; Lopez R, M.C.

    1997-01-01

    The non-ideality of the epithermal neutron flux distribution in a a reactor site parameter (α), the thermal-to-epithermal neutron ratio (f), the irradiation channel neutron temperature (T n ) and the k 0 -factors for more than 20 isotopes were determined in the 3 typical irradiation positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using different experimental methods with conventional and non-conventional monitors. This characterization is used in the k 0 -method of NAA, recently introduced at the Institute. (author). 21 refs., 3 figs., 5 tabs

  18. 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

  19. Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

    International Nuclear Information System (INIS)

    Rice, R.E.

    1976-09-01

    Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments

  20. 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.

  1. 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

  2. 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)

  3. Progress in space nuclear reactor power systems technology development - The SP-100 program

    Science.gov (United States)

    Davis, H. S.

    1984-01-01

    Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

  4. Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods, Stage 18

    International Nuclear Information System (INIS)

    Pazsit, Imre; Nam, Tran Hoai; Dykin, Victor; Jonsson, Anders

    2013-01-01

    This report constitutes Stage 18 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The objective of the research program is to contribute to the strategic research goal of competence and research capacity by building up competence within the Department of Nuclear Engineering at Chalmers University of Technology, regarding reactor physics, reactor dynamics and noise diagnostics. The purpose is also to contribute to the research goal of giving a basis for SSM's supervision by developing methods for identification and localization of perturbations in reactor cores. Results up to Stage 17 were reported in SKI and SSM reports, as listed in the report's summary

  5. Minority and female training programs at the Ford Nuclear Reactor, University of Michigan

    International Nuclear Information System (INIS)

    Burn, R.R.

    1992-01-01

    Nuclear power industry operations staffs are composed predominantly of white males because most of the personnel come from the nuclear submarine and surface branches of the U.S. Navy. The purpose of the minority and female training programs sponsored by the Ford Nuclear Reactor at the University of Michigan is to provide a path for minorities and women to enter the nuclear industry as operators, technicians, and, in the long term, as graduate engineers. The training programs are aimed at high school students, preferably juniors. While the training is directed toward operation of a nuclear reactor, it is equally applicable to careers in most other technical fields. It is hoped that some of the participants will remain at the Ford Nuclear Reactor as reactor operators, enter college, and obtain college degrees, after which they will enter the nuclear industry as graduate engineers

  6. Development of a crack growth analysis is program for reactor head penetration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Yull; Choi, Kwang Hee; Park, Jeong Il [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Kang, Young Hwan; Park, Sung Ho; Kim, Il; Kim, Young Jong; Yoo, Young Joon; Yoo, Wan; Maeng, Wan Young; Choi, Suk Nam; Kim, Kee Suk; Yoon, Sung Won; Kim, Jee Ho; Park, Myung Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-12-31

    Crack growth analysis program for Reactor Head Penetration is being developed for applying to plants such as, Kori 1, Kori 2, Kori 3,4 YoungKwang 1,2 and Uljin 1,2 (1) Stress Evaluation - The stress analysis is required to evaluate the structure integrity for the RVH penetration tubes. The RVH penetration tubes are geometrically non-symmetry except center one. Thus, 3D finite element analysis should be employed for the stress analysis. The magnitude and distribution of residual stress resulted from welding can be determined analytically by simulation welding procedure. (2) Flaw Evaluation - There are two objectives of the penetration tube flaw evaluation to predict the time required for a crack to propagate to the acceptance criteria. The first objective is to perform the parametric evaluation for a postulated crack. The second objective is to develop the flaw evaluation program for the crack detected during the inspection. (3) Characterization of Material Properties of Alloy 600 - These study is to provide data which similarly represent the properties of PWR power plants in Korea. The data is used for analyzing of the stress distribution around penetration tubes. And the PWSCC data will be used for the crack growth rate of the penetration tubes. (author). 92 refs., 121 figs.

  7. Pressurized water reactor nuclear power plant. Environmental characterization information report

    International Nuclear Information System (INIS)

    1981-01-01

    The typical plant chosen for characterization is a 10000-MWe nameplate rating with wet-natural-draft cooling towers and modern radwaste control and processing equipment. The process, plant operating parameters, resources needed, and the environmental residuals and products associated with the power plant are presented. Annual resource usage and pollutant discharges are shown in English and metric units, assuming an annual plant capacity factor of 70%. In addition to annual quantities, the summary table gives quantities in terms of 10 12 Btu (about 293 million kWh) of electrical energy produced for comparison among energy processes. Supporting information and calculation procedures for the data are given. Thirteen environmental points of interest are discussed individually. Cost information, typical radioactive releases, and use of cooling ponds as an alternative cooling method are discussed in appendixes. A glossary and list of acronyms and abbreviations are provided

  8. Nuclear science. U.S. electricity needs and DOE's civilian reactor development program

    International Nuclear Information System (INIS)

    England-Joseph, Judy; Allen, Robert E. Jr.; Fitzgerald, Duane; Young, Edward E. Jr.; Leavens, William P.; Bell, Jacqueline

    1990-05-01

    Electricity projections developed by the North American Electric Reliability Council (NERC) appear to be the best available estimates of future U.S. electricity needs. NERC, which represents all segments of the utility industry, forecasts that before 1998 certain regions of the country, particularly in the more heavily populated eastern half of the United States, may experience shortfalls during summer peak demand periods. These forecasts considered the utility companies' plans, as of 1989, to meet electricity needs during the period; these plans include such measures as constructing additional generators and conducting demand management programs. Working closely with the nuclear industry, DOE is supporting the development of several reactor technologies to ensure that nuclear power remains a viable electricity supply option. In fiscal year 1990, DOE's Civilian Reactor Development Program was funded at $253 million. DOE is using these funds to support industry-led efforts to develop light water reactors (LWR), advanced liquid-metal reactors (LMR), and modular high-temperature gas-cooled reactors (MHTGR) that are safe, environmentally acceptable, and economically competitive. The utility company officials we spoke with, all of whom were in the Southeast, generally supported DOE's efforts in developing these technologies. However, most of the officials do not plan to purchase nuclear reactors until after 2000 because of the high costs of constructing nuclear reactors and current public opposition to nuclear power

  9. The development of the breeder reactor program of the German Federal Republic

    International Nuclear Information System (INIS)

    Haunschild, H.H.

    1984-01-01

    This article recapitulates the stages of the program which is in hand in the German Federal Republic concerning breeder reactors with fast neutrons. In particular, it describes in detail the causes for the delays in the construction of the demonstration reactor SRN 300 (Kalkar) for which the commissioning has now been fixed for 1987. It emphasizes the importance of the collaboration which exists in Europe for the establishment of the projects and the construction of fast neutron power stations, including the SNR 2 reactor; the construction of the latter could begin in the German Federal Republic at the end of the Eighties [fr

  10. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail

  11. Programming of computers for the protection system for Savannah River reactors

    International Nuclear Information System (INIS)

    Finley, R.H.

    1977-06-01

    The monitoring requirements for the SRP Safety Computers are shown. These fast response times coupled with the large number of analog inputs to be scanned imposed stringent program requirements. The system consists of two separate computers, each with its own inputs to monitor half the reactor positions. Either computer can provide the minimum required monitoring. The desired redundant monitoring is provided when both computers are on-line. If both computers are off-line, the reactor is automatically shut down

  12. Reactor R ampersand D programs tough to eliminate; just ask NRC staff

    International Nuclear Information System (INIS)

    Lane, E.

    1993-01-01

    Even if the Clinton administration succeeds in eliminating funding for the advanced liquid metal reactor (ALMR) and modular high-temperature gas reactor (MHTGR) in the fiscal year 1994 budget, it will not wipe out the programs entirely as shown by a recent exchange of letters between the Nuclear Regulatory Commission and the Energy Department. This article examines the political and bureaucratic maneuverings involved in the funding of nuclear power projects

  13. 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

  14. Gas reactor international coope--ative program. Interim report: assessment of gas-cooled reactor economics

    Energy Technology Data Exchange (ETDEWEB)

    1979-08-01

    A computer analysis of domestic economic incentive is presented. Included are the sample computer data set for ten combinations of reprocessing and reactor assumptions; basic data set and computer output; higher uranium availability computer output; 50 percent higher GCR fabrication cost computer output; 50 percent higher GCR reprocessing cost computer output; year 1990 and year 2000 GCR introduction scenario computer outputs; 75 percent perceived capacity factor for PBR computer output; and capital cost of GCRs 1.2 times that of LWRs.

  15. 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

  16. Progress toward acoustic emission characterization for continuous monitoring of reactor pressure vessels

    International Nuclear Information System (INIS)

    Hutton, P.H.; Schwenk, E.B.; Kurtz, R.J.

    1980-01-01

    The purpose of this work is to evaluate the feasibility of detecting and analyzing flaw growth in an operating reactor vessel using acoustic emission (AE) data. A preliminary AE-flaw growth relationship has been developed encompassing six variables. This model relates AE count and energy to fatigue crack driving force ΔK. Pattern recognition has been examined as a means of distinguishing crack growth AE from other acoustic signals with encouraging results. The program is moving into hardware implementation of relationships for testing and refinement by monitoring tests in heavy section specimens and subscale structures. The hardware system will become a reactor monitor prototype. (orig.) [de

  17. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  18. Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

    Directory of Open Access Journals (Sweden)

    Kaiser Krista

    2016-01-01

    Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  19. Synthesis and characterization of nano hydroxyapatite using reverse micro emulsions as nano reactors

    International Nuclear Information System (INIS)

    Amin, S.; Siddique, T.

    2015-01-01

    In the present work reverse micro emulsion has been employed as nano reactors to synthesize nano crystalline Hydroxyapatite (HA). Two precursors; calcium and phosphate with different counter ions of each were used for the synthesis of HA at two different temperatures. To maintain the emulsified nano reactor, cyclohexane, TX-100 and 1-butanol including phosphate precursor was the continuous phase while aqueous Ca precursor solution was taken as the dispersed phase. Nano crystalline particles thus produced were evaluated on the basis of synthesis route, counter ions and temperature. It has been shown that emulsified nano reactors control the morphology, particle size and minimize phase transformation of HA. Characterizations of nano powder of HA are carried out using x-ray diffraction (XRD), Fourier transform infra-red spectroscopy (FTIR), and scanning electron microscopy (SEM). HA crystallite size was found to be in the range of 20-25 nm whereas the morphology of nano particles changed from spheres to rods. (author)

  20. 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)

  1. 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

  2. Closed-loop digital control of nuclear reactors characterized by spatial dynamics

    International Nuclear Information System (INIS)

    Bernard, J.A.; Henry, A.F.; Lanning, D.D.; Meyer, J.E.

    1991-03-01

    This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs

  3. Closed-loop digital control of nuclear reactors characterized by spatial dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Dept. of Nuclear Engineering)

    1991-03-01

    This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs.

  4. 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

  5. Electrical characterization and an equivalent circuit model of a microhollow cathode discharge reactor

    International Nuclear Information System (INIS)

    Taylan, O.; Berberoglu, H.

    2014-01-01

    This paper reports the electrical characterization and an equivalent circuit of a microhollow cathode discharge (MHCD) reactor in the self-pulsing regime. A MHCD reactor was prototyped for air plasma generation, and its current-voltage characteristics were measured experimentally in the self-pulsing regime for applied voltages from 2000 to 3000 V. The reactor was modeled as a capacitor in parallel with a variable resistor. A stray capacitance was also introduced to the circuit model to represent the capacitance of the circuit elements in the experimental setup. The values of the resistor and capacitors were recovered from experimental data, and the proposed circuit model was validated with independent experiments. Experimental data showed that increasing the applied voltage increased the current, self-pulsing frequency and average power consumption of the reactor, while it decreased the peak voltage. The maximum and the minimum voltages obtained using the model were in agreement with the experimental data within 2.5%, whereas the differences between peak current values were less than 1%. At all applied voltages, the equivalent circuit model was able to accurately represent the peak and average power consumption as well as the self-pulsing frequency within the experimental uncertainty. Although the results shown in this paper was for atmospheric air pressures, the proposed equivalent circuit model of the MHCD reactor could be generalized for other gases at different pressures.

  6. Scale-model characterization of flow-induced vibrational response of FFTF reactor internals

    International Nuclear Information System (INIS)

    Ryan, J.A.; Mahoney, J.J.

    1980-10-01

    Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously flow-rate dependent, and the predicted prototype components' response were deemed acceptable

  7. Development, application and also modern condition of the calculated program Imitator of a reactor

    International Nuclear Information System (INIS)

    Aver'yanova, S.P.; Kovel', A.I.; Mamichev, V.V.; Filimonov, P.E.

    2008-01-01

    Features of the calculated program Imitator of a reactor (IR) for WWER-1000 operation simulation are discussed. It is noted that IR application at NPP provides for the project program (BIPR-7) on-line working. This offers a new means, on the one hand, for the efficient prediction and information support of operator, on the other hand, for the verification and development of calculated scheme and neutron-physical model of the WWER-1000 projection program [ru

  8. 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

  9. Safety Evaluation for Packaging for the N Reactor/single pass reactor fuel characterization shipments

    International Nuclear Information System (INIS)

    Stevens, P.F.

    1994-01-01

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the ChemNuclear CNS 1-13G packaging to ship samples of irradiated fuel elements from the 100 K East and 100 K West basins to the Postirradiation Testing Laboratory (PTL) in support of the spent nuclear fuel characterization effort. It also authorizes the return of the fuel element samples to the 100 K East facility using the same packaging. The CNS 1-13G cask has been-chosen to transport the fuel because it has a Certificate of Compliance (CoC) issued by the US Nuclear Regulatory Commission (NRC) for transporting irradiated oxide and metal fuel in commerce. It is capable of being loaded and offloaded underwater and may be shipped with water in the payload compartment

  10. Development of a depletion program for the calculation of the 3D-burn-up dependent power distributions in light water reactors

    International Nuclear Information System (INIS)

    Bennewitz, F.; Mueller, A.; Wagner, M.R.

    1977-11-01

    Based on the nodal collision probability method a multi-dimensional reactor burn-up program MEDIUM has been developed, which is written for 2 neutron energy groups. It is characterized by high computing speed, considerable generality and flexibility, a number of useful program options and good accuracy. The three-dimensional flux calculation model is described, the formulation and method of solution of the nuclear depletion equations and further details of the program structure. The results of a number of comparisons with experimental data and with independent computer programs are presented. (orig.) [de

  11. University Reactor Instrumentation grant program. Final report, September 7, 1990--August 31, 1995

    International Nuclear Information System (INIS)

    Talnagi, J.W.

    1998-01-01

    The Ohio State University Nuclear Reactor Laboratory (OSU NRL) participated in the Department of Energy (DOE) grant program commonly denoted as the University Reactor Instrumentation (URI) program from the period September 1990 through August 1995, after which funding was terminated on a programmatic basis by DOE. This program provided funding support for acquisition of capital equipment targeted for facility upgrades and improvements, including modernizing reactor systems and instrumentation, improvements in research and instructional capabilities, and infrastructure enhancements. The staff of the OSU NRL submitted five grant applications during this period, all of which were funded either partially or in their entirety. This report will provide an overview of the activities carried out under these grants and assess their impact on the OSU NRL facilities

  12. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1982-August 1983

    International Nuclear Information System (INIS)

    Brenizer, J.S.; Benneche, P.E.

    1984-03-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1982 was over 5500 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1982 through August 1983

  13. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1983-August 1984

    International Nuclear Information System (INIS)

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

    1984-11-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1983 was over 6000 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1983 through August 1984

  14. 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

  15. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    Energy Technology Data Exchange (ETDEWEB)

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  16. Immediate relation of ING to fast breeder reactor programs

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, W B

    1969-07-01

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  17. Immediate relation of ING to fast breeder reactor programs

    International Nuclear Information System (INIS)

    Lewis, W.B.

    1969-01-01

    The future large-scale use of nuclear energy is linked in the United States and other major countries to their fast breeder reactor development. Very serious basic problems have been discovered within the last two years, limiting the life in the high fast neutron flux at appropriate temperatures of materials, in particular of metals suitable for fuel cladding in sodium coolant. There is therefore a most urgent need for materials testing facilities under controlled conditions of temperature and neutron flux at sufficiently high ratings to match or surpass those required in commercially competitive fast breeder reactors. None of the test facilities yet planned for 1976 or sooner in the western world appears to match these conditions. The problem is mainly the difficulty of providing the high neutron flux effectively continuously. The spallation reaction in heavy elements was chosen as the basis of ING - the intense neutron generator, because it is the only known reaction that promises a fast neutron source density that is higher than can be controlled from the fission process. It is suggested that several countries will wish to consider urgently whether they should also explore the spallation reaction for the purpose of a fast neutron irradiation test facility. In view of the discontinuance of the ING project in Canada a favourable opportunity will exist over the next few months 10 obtain from Canada by direct personal contact details of the significant study that has been carried on for ING over the last five years. In the event that satisfactory materials are established within the lifetime of the spallation facilities they may continue to be used for the production of selected isotopes more profitably produced in high neutron fluxes. The facilities may be also used for the desirable preirradiation of thorium reactor fuel. The other research purposes planned for ING could also be served. (author)

  18. The Thermos program for nuclear reactors specialized in district heating

    International Nuclear Information System (INIS)

    Lerouge, B.

    1976-01-01

    Many studies have been made in France on the use of nuclear heat for district heating. After a brief account of the problems raised by the use of thermal waste from big nuclear power stations, the quantitative and qualitative needs of heating networks are analyzed and the Thermos project described. This is a very robust reactor of the pool type, with an output of 100MW, supplying low-pressure water at 100 deg C. The advantages from the aspects of safety and economy are described, and the present state of the project and its possible developments summarized [fr

  19. New approaches in the USA for high-temperature gas-cooled reactors. Gas-Cooled Reactor Programs

    International Nuclear Information System (INIS)

    Kasten, P.R.; Neylan, A.J.; Penfield, S.R. Jr.

    1985-08-01

    Several concepts are being evaluated in the US HTR Program to explore designs which might improve the commercial viability of nuclear power. The general approach is to reduce the reactor power and increase the ability to use inherent features for removing heat following extreme accidents. The unit size and design of these concepts are constrained so that extreme accidents do not result in significant release of radioactivity from the reactor plant. Through the greater reliance on inherent safety features in small HTRs, it should be possible to minimize the amount of nuclear grade components required in the balance-of-plant, which could lead to an economic system. Four HTR concepts are presently being evaluated within the US Program, and these concepts are briefly summarized. A modular HTR using a steel pressure vessel, which is very similar to one of the four HTR concepts being evaluated within the US National program, is presented as an example of a specific concept to illustrate the features and performance of HTRs having a high degree of inherent safety

  20. Fast reactor test facilities in the US safety program

    International Nuclear Information System (INIS)

    Avery, R.; Dickerman, C.E.; Lennox, D.H.; Rose, D.

    1979-01-01

    The needs for safety information derivable from in-pile programs are reviewed, and the correlation made with existing and planned capability. In view of the current status of the U.S. breeder program, emphasis is given in the review to the impact of different fast breeder options on the required program and facilities. It is concluded that facility needs are somewhat independent of specific fast breeder concept, even though the relative emphasis on the various safety issues will differ. 8 refs

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  2. University Reactor Sharing Program. Period covered: September 1, 1981-August 31, 1982

    International Nuclear Information System (INIS)

    Hajek, B.K.; Myser, R.D.; Miller, D.W.

    1982-12-01

    During the period from September 1, 1981 to August 31, 1982, the Ohio State University Nuclear Reactor Laboratory participated in the Reactor Sharing Program by providing services to eight colleges and universities. A laboratory on Neutron Activation Analysis was developed for students in the program. A summary of services provided and a copy of the laboratory procedure are attached. Services provided in the last funded period were in three major areas. These were neutron activation analysis, nuclear engineering labs, and introductions to nuclear research. One group also performed radiation surveys and produced isotopes for calibration of their own analytical equipment

  3. 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

  4. Development programs on decommissioning technology for reactors and fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Fujiki, K.

    1992-01-01

    The Science and Technology Agency (STA) of Japan is promoting technology development for decommissioning of nuclear facilities by entrusting various research programs to concerned research organisations: JAERI, PNC and RANDEC, including first full scale reactor decommissioning of JPDR. According to the results of these programs, significant improvement on dismantling techniques, decontamination, measurement etc. has been achieved. Further development of advanced decommissioning technology has been started in order to achieve reduction of duration of decommissioning work and occupational exposures in consideration of the decommissioning of reactors and fuel cycle facilities. (author) 5 refs.; 7 figs.; 1 tab

  5. 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)

  6. 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

  7. Resolving Nuclear Reactor Lifetime Extension Questions: A Combined Multiscale Modeling and Positron Characterization approach

    International Nuclear Information System (INIS)

    Wirth, B; Asoka-Kumar, P; Denison, A; Glade, S; Howell, R; Marian, J; Odette, G; Sterne, P

    2004-01-01

    The objective of this work is to determine the chemical composition of nanometer precipitates responsible for irradiation hardening and embrittlement of reactor pressure vessel steels, which threaten to limit the operating lifetime of nuclear power plants worldwide. The scientific approach incorporates computational multiscale modeling of radiation damage and microstructural evolution in Fe-Cu-Ni-Mn alloys, and experimental characterization by positron annihilation spectroscopy and small angle neutron scattering. The modeling and experimental results are

  8. Research program and uses of the solution fueled reactor SILENE

    International Nuclear Information System (INIS)

    Barbry, F.; Ratel, R.

    1985-09-01

    Designed and operated by the Nuclear Protection and Safety Institute of the CEA, SILENE is an original small sized reactor fueled with an uranyl nitrate solution. The reactor is capable to operate in three modes: ''Pulse'' operation (high power levels up to 1000 Megawatts during several millisecond), ''Free evolution'' operation (simulation of criticality accident excursions), ''Steady state'' operation in a power range of 0.01 W to 1 kW. The core can be surrounded by appropriate shields (lead, polyethylene) to vary the leakage radiations and the gamma to neutron dose ratio. It's possible to insert in the central cavity of the annular core vessel some capsules, devices or samples to be submitted to very high radiations levels. The research activities are mainly devoted towards nuclear safety studies: the criticality accident studies, and the behavior of oxide fuels under transient conditions. Some examples of tests are presented. As to other applications of the SILENE facility, the main studies now in progress deal with: designing and calibration of Health physics intrumentation, neutron and gamma dosimetry, and, radiobiology. Once the characteristics of radiation field are qualified by calculations and experimental techniques, SILENE will be proposed as a reference source [fr

  9. 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

  10. 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)

  11. Optimization of fuel management and control poison of a nuclear power reactor by dynamic programming

    International Nuclear Information System (INIS)

    Lima, C.A.R. de.

    1977-01-01

    The distribution of fuel and control poison in a nuclear reactor was optimized by the method of Dynamic Programming. A 620 M We Pressurized Water Reactor similar to Angra-1 was studied. The reactor operation was simulated in a IBM-1130 computer. Two fuel shuffling schemes and three poison management schemes were simultaneously employed in the reactor divided into three regions of equal volume and two consecutive stages were studied in order to determine the influence of poison management on the optimum fuel management policy. When uniform poisoning on all the three regions was permitted the traditional out-in fuel management policy proved to be more economic. On introducing simultaneous poison management, the optimum fuel management sequence was found to be different. The results obtained indicate a stronger interaction between the fuel management and the poison management than anticipated in previous works. (author)

  12. Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Mejia C, M. A.

    2016-01-01

    The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

  13. 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)

  14. 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

  15. 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.

  16. In situ TEM and synchrotron characterization of U–10Mo thin specimen annealed at the fast reactor temperature regime

    International Nuclear Information System (INIS)

    Yun, Di; Mo, Kun; Mohamed, Walid; Ye, Bei; Kirk, Marquis A.; Baldo, Peter; Xu, Ruqing; Yacout, Abdellatif M.

    2015-01-01

    U–Mo metallic alloys have been extensively used for the Reduced Enrichment for Research and Test Reactors (RERTR) program, which is now known as the Office of Material Management and Minimization under the Conversion Program. This fuel form has also recently been proposed as fast reactor metallic fuels in the recent DOE Ultra-high Burnup Fast Reactor project. In order to better understand the behavior of U–10Mo fuels within the fast reactor temperature regime, a series of annealing and characterization experiments have been performed. Annealing experiments were performed in situ at the Intermediate Voltage Electron Microscope (IVEM-Tandem) facility at Argonne National Laboratory (ANL). An electro-polished U–10Mo alloy fuel specimen was annealed in situ up to 700 °C. At an elevated temperature of about 540 °C, the U–10Mo specimen underwent a relatively slow microstructure transition. Nano-sized grains were observed to emerge near the surface. At the end temperature of 700 °C, the near-surface microstructure had evolved to a nano-crystalline state. In order to clarify the nature of the observed microstructure, Laue diffraction and powder diffraction experiments were carried out at beam line 34-ID of the Advanced Photon Source (APS) at ANL. Phases present in the as-annealed specimen were identified with both Laue diffraction and powder diffraction techniques. The U–10Mo was found to recrystallize due to thermally-induced recrystallization driven by a high density of pre-existing dislocations. A separate in situ annealing experiment was carried out with a Focused Ion Beam processed (FIB) specimen. A similar microstructure transition occurred at a lower temperature of about 460 °C with a much faster transition rate compared to the electro-polished specimen. - Highlights: • TEM annealing experiments were performed in situ at the IVEM facility up to fast reactor temperature. • At 540 °C, the U-10Mo specimen underwent a slow microstructure transition

  17. In situ TEM and synchrotron characterization of U–10Mo thin specimen annealed at the fast reactor temperature regime

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Di, E-mail: diyun1979@xjtu.edu.cn [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Xi' an Jiao Tong University, 28 Xian Ning West Road, Xi' an 710049 (China); Mo, Kun; Mohamed, Walid; Ye, Bei; Kirk, Marquis A.; Baldo, Peter; Xu, Ruqing; Yacout, Abdellatif M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2015-12-15

    U–Mo metallic alloys have been extensively used for the Reduced Enrichment for Research and Test Reactors (RERTR) program, which is now known as the Office of Material Management and Minimization under the Conversion Program. This fuel form has also recently been proposed as fast reactor metallic fuels in the recent DOE Ultra-high Burnup Fast Reactor project. In order to better understand the behavior of U–10Mo fuels within the fast reactor temperature regime, a series of annealing and characterization experiments have been performed. Annealing experiments were performed in situ at the Intermediate Voltage Electron Microscope (IVEM-Tandem) facility at Argonne National Laboratory (ANL). An electro-polished U–10Mo alloy fuel specimen was annealed in situ up to 700 °C. At an elevated temperature of about 540 °C, the U–10Mo specimen underwent a relatively slow microstructure transition. Nano-sized grains were observed to emerge near the surface. At the end temperature of 700 °C, the near-surface microstructure had evolved to a nano-crystalline state. In order to clarify the nature of the observed microstructure, Laue diffraction and powder diffraction experiments were carried out at beam line 34-ID of the Advanced Photon Source (APS) at ANL. Phases present in the as-annealed specimen were identified with both Laue diffraction and powder diffraction techniques. The U–10Mo was found to recrystallize due to thermally-induced recrystallization driven by a high density of pre-existing dislocations. A separate in situ annealing experiment was carried out with a Focused Ion Beam processed (FIB) specimen. A similar microstructure transition occurred at a lower temperature of about 460 °C with a much faster transition rate compared to the electro-polished specimen. - Highlights: • TEM annealing experiments were performed in situ at the IVEM facility up to fast reactor temperature. • At 540 °C, the U-10Mo specimen underwent a slow microstructure transition

  18. 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)

  19. 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)

  20. 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

  1. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  2. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1981-August 1982

    International Nuclear Information System (INIS)

    Brenizer, J.S.; Benneche, P.E.

    1982-12-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1981 and nearly 5000 megawatt-hours) research reactor in the mid-Atlantic States. In addition, a second, small (50 watt) reactor is also available for use in educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1981 through August 1982

  3. A mathematical method for boiling water reactor control rod programming

    International Nuclear Information System (INIS)

    Tokumasu, S.; Hiranuma, H.; Ozawa, M.; Yokomi, M.

    1985-01-01

    A new mathematical programming method has been developed and utilized in OPROD, an existing computer code for automatic generation of control rod programs as an alternative inner-loop routine for the method of approximate programming. The new routine is constructed of a dual feasible direction algorithm, and consists essentially of two stages of iterative optimization procedures Optimization Procedures I and II. Both follow almost the same algorithm; Optimization Procedure I searches for feasible solutions and Optimization Procedure II optimizes the objective function. Optimization theory and computer simulations have demonstrated that the new routine could find optimum solutions, even if deteriorated initial control rod patterns were given

  4. Reactor operations, inspection and maintenance. PNGS Calibration Program

    International Nuclear Information System (INIS)

    Lopez, E.

    1997-01-01

    The PNGS Calibration Program is being implemented as a response to various concerns identified in recent PEER evaluations and AECB audits. Identified areas of concern were the approach to instrument calibration of Special Safety Systems (SSS). The implementation of a calibration program is a significant improvement in operating practices. A systematic and comprehensive approach to calibration of instrumentation will improve the quality of operation of the plant with a positive contribution to PNGS safety of operation and economic objectives. This paper describes the strategy to implement the proposed calibration program and describes its calibration data requirements. (DM)

  5. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  6. Fast reactor development program in France in 1999

    International Nuclear Information System (INIS)

    Leclere, Jacques; Teilland, C.; Astegiano, J.C.; Desreumaux, J.; Grenet, P.; Del Beccaro, R.; Vasile, A.

    2000-01-01

    The year 1999 was marked by important industrial reorganisations, especially concerning Framatome: the shareholding was modified, and at the end of the year the fusion of nuclear activities with those of Siemens was decided. The construction of the last N4 plants is completed, with no realisation expected in the near future; the decision on an EPR unit is not expected in the short term. The operation of the plants was satisfying, the availability is nevertheless lower than in 1998. Renovation, inspection and maintenance works were actively conducted at Phenix; decommissioning work was undertaken at Superphenix; the core unloading began in December. The R and D strategy was redefined; on fast reactors, work decreases on sodium and increases on other variants. International collaborations remain active. (author)

  7. Gas reactor international cooperative program. HTR-synfuel application assessment

    International Nuclear Information System (INIS)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H 2 +CO 2 ) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000 0 F steam is generated at the industrial user sites. The products of methanation (CH 4 + H 2 O) are piped back to the reformer at the central station HTR

  8. Gas reactor international cooperative program. HTR-synfuel application assessment

    Energy Technology Data Exchange (ETDEWEB)

    1979-09-01

    This study assesses the technical, environmental and economic factors affecting the application of the High Temperature Gas-Cooled Thermal Reactor (HTR) to: synthetic fuel production; and displacement of fossil fuels in other industrial and chemical processes. Synthetic fuel application considered include coal gasification, direct coal liquefaction, oil shale processing, and the upgrading of syncrude to motor fuel. A wide range of other industrial heat applications was also considered, with emphasis on the use of the closed-loop thermochemical energy pipeline to supply heat to dispersed industrial users. In this application syngas (H/sub 2/ +CO/sub 2/) is produced at the central station HTR by steam reforming and the gas is piped to individual methanators where typically 1000/sup 0/F steam is generated at the industrial user sites. The products of methanation (CH/sub 4/ + H/sub 2/O) are piped back to the reformer at the central station HTR.

  9. A spare-parts inventory program for TRIGA reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, T V; Ringle, J C; Johnson, A G [Oregon State University (United States)

    1974-07-01

    As is fairly common with new reactor facilities, we had a few spare parts on hand as part of our original purchase when the OSU TRIGA first went critical in March of 1967. Within a year or so, however, it became apparent that we should critically examine our spare parts inventory in order to avoid unnecessary or prolonged outages due to lack of a crucial piece of equipment. Many critical components (those which must be present and operable according to our license or technical specifications) were considered, and a priority list of acquiring these was established. This first list was drawn up in March, 1969, two years after initial criticality, and some key components were ordered. The availability of funds was the overriding restriction then and now. This spare-parts list is reviewed and new components purchased annually; the average amount spent has been about $2,000 per year. This inventory has proved invaluable more than once; without it, we would have had lengthy shutdowns awaiting the arrival of the needed component. The sobering thought, however, is that our spare-parts inventory is still not complete-far from it, in fact, because this would be prohibitively expensive. It is very difficult to guess with 100% accuracy just which component might need replacing, and your $10,000 inventory of spare parts is useless in that instance if it doesn't include the needed part. An idea worth considering is to either (a) encourage General Atomic, through the collective voice of all TRIGA owners, to maintain a rather complete inventory of replacement parts, or (b) maintain an owner's spare-parts pool, financed by contributions from all the facilities. If either of these pools was established, the needed part could reach any facility within the U.S. within a few days, minimizing reactor outage time. (author)

  10. A spare-parts inventory program for TRIGA reactors

    International Nuclear Information System (INIS)

    Anderson, T.V.; Ringle, J.C.; Johnson, A.G.

    1974-01-01

    As is fairly common with new reactor facilities, we had a few spare parts on hand as part of our original purchase when the OSU TRIGA first went critical in March of 1967. Within a year or so, however, it became apparent that we should critically examine our spare parts inventory in order to avoid unnecessary or prolonged outages due to lack of a crucial piece of equipment. Many critical components (those which must be present and operable according to our license or technical specifications) were considered, and a priority list of acquiring these was established. This first list was drawn up in March, 1969, two years after initial criticality, and some key components were ordered. The availability of funds was the overriding restriction then and now. This spare-parts list is reviewed and new components purchased annually; the average amount spent has been about $2,000 per year. This inventory has proved invaluable more than once; without it, we would have had lengthy shutdowns awaiting the arrival of the needed component. The sobering thought, however, is that our spare-parts inventory is still not complete-far from it, in fact, because this would be prohibitively expensive. It is very difficult to guess with 100% accuracy just which component might need replacing, and your $10,000 inventory of spare parts is useless in that instance if it doesn't include the needed part. An idea worth considering is to either (a) encourage General Atomic, through the collective voice of all TRIGA owners, to maintain a rather complete inventory of replacement parts, or (b) maintain an owner's spare-parts pool, financed by contributions from all the facilities. If either of these pools was established, the needed part could reach any facility within the U.S. within a few days, minimizing reactor outage time. (author)

  11. Fast reactor safety program. Progress report, January-March 1980

    International Nuclear Information System (INIS)

    1980-05-01

    The goal of the DOE LMFBR Safety Program is to provide a technology base fully responsive to safety considerations in the design, evaluation, licensing, and economic optimization of LMFBRs for electrical power generation. A strategy is presented that divides safety technology development into seven program elements, which have been used as the basis for the Work Breakdown Structure (WBS) for the Program. These elements include four lines of assurance (LOAs) involving core-related safety considerations, an element supporting non-core-related plant safety considerations, a safety R and D integration element, and an element for the development of test facilities and equipment to be used in Program experiments: LOA-1 (prevent accidents); LOA-2 (limit core damage); LOA-3 (maintain containment integrity); LOA-4 (attenuate radiological consequences); plant considerations; R and D integration; and facility development

  12. The Alternate Technology Program for Aluminum Research Reactor Spent Fuel

    International Nuclear Information System (INIS)

    Barlow, M.W.

    1998-01-01

    This paper describes the program for disposition of aluminum-based RRSNF, including the requirements for road-ready dry storage and repository disposal and the criteria to be considered in selecting among the alternative technologies

  13. 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.

  14. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

  15. Annual report in compliance with the reactor sharing program, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    Karam, R.A.

    1997-01-01

    This report contains information with regard to facilities utilization, descriptions (brief), personnel, organization, and programs of the Neely Nuclear Research Center (NNRC) at the Georgia Institute of Technology. The NNRC has two major facilities: the Georgia Tech Research Reactor and the Hot Cell Laboratory. This report of NNRC utilization is prepared in compliance with the contract requirements between the U.S. Department of Energy and the Georgia Institute of Technology. The NNRC is a participant in the University Reactor Sharing Program; as such, it makes available its 5 MW research reactor, its Co-60 irradiation facility and its activation analysis laboratory to large numbers of students and faculty from many universities and colleges

  16. Area Safety Program for the tokamak fusion test reactor (TFTR)

    International Nuclear Information System (INIS)

    Rappe, G.M.

    1984-10-01

    Overall the Area Safety Program has proved to be a very successful operation. There is no doubt that a safety program organized through line management is the best way to involve all personnel. Naturally, when the program was first started, there was some criticism and a certain resistance on the part of a few individuals to fully participate. However, once the program was underway and it could be seen that it was working to everyone's advantage, this reluctance disappeared and a spirit of full cooperation is now enjoyed. It is very important that for this success to continue there must be a two way flow of information, both from the Area Safety Coordinators up through line management, and from senior management, with decisions and answers, back down through the management chain with the utmost dispatch. As with all programs, there is still room for improvement. This program has started a review cycle with a view to streamlining certain areas and possibly increasing its scope in others

  17. The role of a technology demonstration program for future reactors

    International Nuclear Information System (INIS)

    Viktorov, A.

    2011-01-01

    A comprehensive technology demonstration program is seen as an important component of the overall safety case, especially for a novel technology. The objective of such a program is defined as providing objective and auditable evidence that the technology will meet or exceed the relevant requirements. Various aspects of such a program are identified and then discussed in some details in this presentation. We will show how the need for such a program is anchored in fundamental safety principles. Attributes of the program, means of achieving its objective, roles of participants, as well as key steps are all elaborated. It will be argued that to prove a novel technology, the designer will have to combine several activities such as the use of operational experience, prototyping of the technology elements, conduct of experiments and tests under representative conditions, as well as modeling and analysis. Importance of availability of experimental facilities and qualified scientific and technical staff is emphasized. A solid technology demonstration program will facilitate and speed up regulatory evaluations of licensing applications. (author)

  18. The role of a technology demonstration program for future reactors

    Energy Technology Data Exchange (ETDEWEB)

    Viktorov, A. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2011-07-01

    A comprehensive technology demonstration program is seen as an important component of the overall safety case, especially for a novel technology. The objective of such a program is defined as providing objective and auditable evidence that the technology will meet or exceed the relevant requirements. Various aspects of such a program are identified and then discussed in some details in this presentation. We will show how the need for such a program is anchored in fundamental safety principles. Attributes of the program, means of achieving its objective, roles of participants, as well as key steps are all elaborated. It will be argued that to prove a novel technology, the designer will have to combine several activities such as the use of operational experience, prototyping of the technology elements, conduct of experiments and tests under representative conditions, as well as modeling and analysis. Importance of availability of experimental facilities and qualified scientific and technical staff is emphasized. A solid technology demonstration program will facilitate and speed up regulatory evaluations of licensing applications. (author)

  19. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  20. Preparation, Characterization and Adsorption Evaluation of old Newspaper Fibres using Basket Reactor (Nickel Removal by Adsorption)

    International Nuclear Information System (INIS)

    Ossman, M. E.; Abdelfatah, M.; Kiros, Y.

    2016-01-01

    In this work, old newspaper fibers bleached with H 2 O 2 , treated with KOH and treated with NaOCl were investigated as potential adsorbents. The characterization of the produced fibers using FTIR, SEM and particle size distribution have been carried out and tested for the removal of Ni (II) from aqueous solutions. The results indicated that the fibers treated with KOH give the highest %removal of Ni (II) with 88%. Two different reactor designs (batch and basket reactor) with different variables were studied. The results indicated that the equilibrium time was 30 min. and the removal of Ni (II) increased significantly as the p H increased from 2.0 to 6.0 and decreased at p H range of 6.5–8.0. The adsorption of Ni (II) onto old newspaper fibers treated with KOH using batch and basket reactors follows the Langmuir isotherm. The pseudo second order kinetic model provided good correlation for the adsorption of Ni (II) onto old newspaper fibers treated with KOH for both batch and basket reactors.

  1. Control program of the neutron four-circle-diffractometer P32 at the SILOE reactor/CEN Grenoble

    International Nuclear Information System (INIS)

    Guth, H.; Paulus, H.; Reimers, W.; Heger, G.

    1983-09-01

    The four-circle diffractometer P32 for elastic neutron scattering on single crystals was installed at the SILOE reactor/CEN Grenoble in 1981. The control program, presented here, is a new update of the former program versions used at the FR2 reactor/Kernforschungszentrum Karlsruhe. Important improvements concerning reliability and handling of the diffractometer are added. (orig.) [de

  2. 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...

  3. 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.

  4. 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

  5. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    International Nuclear Information System (INIS)

    Bissani, M; O'Kelly, D S

    2006-01-01

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

  6. Multimodal shipments under program on Russian-origin research reactor SFA return to Russian Federation

    International Nuclear Information System (INIS)

    Dorofeev, A.N.; Ivashchenko, A.A.; Kanashov, B.A.; Komarov, S.V.; Komarov, S.N.; Barinkov, O.P.

    2010-01-01

    The paper describes experience in preparation and organization of research reactor nuclear material import under the Program on Russian-Origin Highly Enriched Uranium Return to the Russian Federation. It also summarizes evolution of transport equipment, conveyances and routes and describes types of packages, their adaptation and certification, safety issues, peculiarities and prospective use of the packagings and conveyances. (author)

  7. Program system for calculating streaming neutron radiation field in reactor cavity

    International Nuclear Information System (INIS)

    He Zhongliang; Zhao Shu.

    1986-01-01

    The A23 neutron albedo data base based on Monte Carlo method well agrees with SAIL albedo data base. RSCAM program system, using Monte Carlo method with albedo approach, is used to calculate streaming neutron radiation field in reactor cavity and containment operating hall. The dose rate distributions calculated with RSCAM in square concrete duct well agree with experiments

  8. 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

  9. 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

  10. 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)

  11. Overview of AEOD's program for trending reactor operational events

    International Nuclear Information System (INIS)

    Baranowsky, P.W.; O'Reilly, P.D.; Rasmuson, D.M.; Houghton, J.R.

    1994-01-01

    This paper presents an overview of the trending program being performed by AEOD. The major elements of the program include: (1) system and component reliability trending and analysis, (2) special data collection and analysis (e.g., IPE and PRA component failure data, common cause failure event data), (3) risk assessment of safety issues based on actual operating experience, (4) Accident Sequence Precursor (ASP) Program, and (5) trending US industry risk. AEOD plans to maintain up-to-date safety data trends for selected high risk or high regulatory profile components, systems, accident initiators, accident sequences, and regulatory issues. AEOD will also make greater use of PRA insights and perform limited probabilistic safety assessments to evaluate the safety significance of qualitative results. Examples of a system study and an issue evaluation are presented, as well as a summary of the common cause failure event database

  12. Observations concerning a new DOE Civilian Reactor Development Program derived from the NPOVS task

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1985-01-01

    The Nuclear Power Options Viability Study (NPOVS) addressed many issues which relate to a DOE program plan. The study, by design, did not compare reactor types; however, it provides extensive background which can be drawn upon for the purposes of planning a new program. The discussion and recommendations of this paper draw heavily on NPOVS but are not constrained to the NPOVS time frame or to its emphasis on passive safety

  13. Status and future program of reactor physics experiments in JAERI Critical facilities, FCA and TCA

    International Nuclear Information System (INIS)

    Okajima, Shigeaki; Osugi, Toshitaka; Nakajima, Ken; Suzaki, Takenori; Miyoshi, Yoshinori

    1999-01-01

    The critical facilities in JAERI, FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly), have been used to provide integral data for evaluation of nuclear data as well as for development of various types of reactor since they went critical in 1960's. In this paper a review is presented on the experimental programs in both facilities. And the experimental programs in next 5 years are also shown. (author)

  14. Characterization of aerosols from industrial fabrication of mixed-oxide nuclear reactor fuels

    International Nuclear Information System (INIS)

    Hoover, M.D.; Newton, G.J.

    1997-01-01

    Recycling plutonium into mixed-oxide (MOX) fuel for nuclear reactors is being given serious consideration as a safe and environmentally sound method of managing plutonium from weapons programs. Planning for the proper design and safe operation of the MOX fuel fabrication facilities can take advantage of studies done in the 1970s, when recycling of plutonium from nuclear fuel was under serious consideration. At that time, it was recognized that the recycle of plutonium and uranium in irradiated fuel could provide a significant energy source and that the use of 239 Pu in light water reactor fuel would reduce the requirements for enriched 235 U as a reactor fuel. It was also recognized that the fabrication of uranium and plutonium reactor fuels would not be risk-free. Despite engineered safety precautions such as the handling of uranium and plutonium in glove-box enclosures, accidental releases of radioactive aerosols from normal containment might occur. Workers might then be exposed to the released materials by inhalation

  15. Moving into the 21st century - The United States' Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Reilly, Jill E.

    1999-01-01

    Since 1996, when the United States Department of Energy and the Department of State jointly adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, twelve shipments totaling 2,985 MTR and TRIGA spent nuclear fuel assemblies from research reactors around the world have been accepted into the United States. These shipments have contained approximately 1.7 metric tons of HEU and 0.6 metric tons of LEU. Foreign research reactor operators played a significant role in this success. A new milestone in the acceptance program occurred during the summer of 1999 with the arrival of TRIGA spent nuclear fuel from Europe through the Charleston Naval Weapons Station via the Savannah River Site to the Idaho National Engineering and Environmental Laboratory. This shipment consisted of five casks of TRIGA spent nuclear fuel from research reactors in Germany, Italy, Slovenia, and Romania. These casks were transported by truck approximately 2,400 miles across the United States (one cask packaged in an ISO container per truck). Drawing upon lessons learned in previous shipments, significant technical, legal, and political challenges were addressed to complete this cross-country shipment. Other program activities since the last RERTR meeting have included: formulation of a methodology to determine the quantity of spent nuclear fuel in a damaged condition that may be transported in a particular cask (containment analysis for transportation casks); publication of clarification of the fee policy; and continued planning for the outyears of the acceptance policy including review of reactors and eligible material quantities. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues to demonstrate success due to the continuing commitment between the United States and the research reactor community to make this program work. We strongly encourage all eligible research reactors to decide as soon as possible to

  16. 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)

  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. Overview of the US program of controls for advanced reactors

    International Nuclear Information System (INIS)

    White, J.D.; Sackett, J.I.; Monson, R.; Lindsay, R.W.; Carroll, D.G.

    1989-01-01

    An automated control system can incorporate control goals and strategies, assessment of present and future plant status, diagnostic evaluation and maintenance planning, and signal and command validation. It has not been feasible to employ these capabilities in conventional hard-wired, analog, control systems. Recent advances in computer-based digital data acquisition systems, process controllers, fiber-optic signal transmission artificial intelligence tools and methods, and small inexpensive, fast, large-capacity computers---with both numeric and symbolic capabilities---have provided many of the necessary ingredients for developing large, practical automated control systems. Furthermore, recent reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. This paper presents an overall US national perspective for advanced controls research and development. The goals of high reliability, low operating cost and simple operation are described. The staged approach from conceptualization through implementation is discussed. Then the paper describes the work being done by ORNL, ANL and GE. The relationship of this work to the US commercial industry is also discussed

  19. A review of fast reactor program in Japan. April 2000 - March 2001

    International Nuclear Information System (INIS)

    Nagata, Takashi; Yamashita, Hidetoshi

    2001-01-01

    This report describes the development and activities on fast reactors in Japan thru April 2000 to March 2001. During this period, the most important result of the Japanese Fast Reactor Project was that the first phase 'Feasibility Study on Commercialized Fast Reactor Cycle Systems' was completed at the end of March 2001, and the second phase study has just started in order to narrow down the candidate concepts selected in the first phase for next stage. In the Experimental Fast Reactor 'Joyo', the 35 th rated power operation was completed by the end of May 2000. The 13 th periodical inspection and reconstruction works for the Joyo upgrading program (MK-III) were started on the beginning of June 2000. The modification of the cooling system is underway. In the Prototype Fast Breeder Reactor 'Monju', countermeasures against sodium leakage have already been drawn up based on 'Monju' comprehensive safety review. The Japan Atomic Energy Commission (JAEC) has issued a new 'Long-term Program for Research, Development and Utilization of Nuclear Energy' in November 2000. (author)

  20. Technology programs in support of advanced light water reactors

    International Nuclear Information System (INIS)

    1990-03-01

    Design for Constructability's overall purpose was to identify and address changes in the nuclear industry to restore nuclear energy as an attractive option for new generating capacity. The program stove to meet the following goals related to the future construction of nuclear power plants: reduced costs; assurance of improved quality; and shortened construction schedules. This is Volume 1 of three volumes

  1. Nuclear reactor structure materials study (Survey of relevant ISTC programs)

    Energy Technology Data Exchange (ETDEWEB)

    Tocheny, L.V.; Godowski, W. (ISTC-International Science and Technology Center, Moscow (Russian Federation)); Ballesteros, A. (Tecnatom S.A., Materials and Life Management, Madrid (Spain)), e-mail: aballesteros@tecnatom.es; Deffrennes, M.; Hugon, M. (European Commission, DG RTD, Brussels (Belgium))

    2009-07-01

    The international co-operation and programs of ISTC (The International Science and Technology Center) are reviewed in the fields covered by the SMIRT20 Programme. ISTC was created in Moscow in 1994 by Russia, EU, USA, and Japan. Presently, ISTC has 40 member countries (including 27 Member States of the EU), representing the CIS, Europe, Asia, and North America

  2. Reactor physical program in the frame of the MSR-SPHINX transmuter concept development

    International Nuclear Information System (INIS)

    Hron, M.; Mikisek, M.

    2008-01-01

    In the frame of the R and D program for the Molten Salt Reactor (MSR) - SPHINX (Spent Hot fuel Incinerator by Neutron flux) concept, which has been under development in the Czech Republic as an actinide burner in resonance neutron spectrum and a radionuclide transmuter in a well-thermalized neutron spectrum, and namely its reactor physical part, the relatively broad experimental activities have been involved in the program, recently, which will serve for a validation of computer codes and verification of design inputs for designing of a demonstration unit of the MSR-type. The experimental program, which 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, will be in the proposed next stage of the program focused on a large-scale experimental verification of design inputs by use of MSR-type inserted zones into the existing light water moderated experimental reactor LR-0, which may allow to modify it to experimental zero power salt reactor SR-0. There has been a preparatory stage of the project called EROS started in the year 2006 and new experiments with MSR-type zones irradiated by cyclotron based neutron source are planned at the end of 2007 and should go on in the year 2008. There will be a brief description of the so far prepared and performed experimental programs introduced in the paper. (authors)

  3. 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

  4. Measurements of combined neutron and photon fluxes for the accurate characterization of the future Jules Horowitz irradiation reactor experimental conditions

    International Nuclear Information System (INIS)

    Fourmentel, D.

    2013-01-01

    A new Material Testing Reactor (MTR), the Jules Horowitz Reactor (JHR), is under construction at the CEA Cadarache (French Alternatives Energies and Atomic Energy Commission). From 2016 this new MTR will be a new facility for the nuclear research on materials and fuels. The quality of the experiments to be conducted in this reactor is largely linked to the good knowledge of the irradiation conditions. Since 2009, a new research program called IN-CORE1 'Instrumentation for Nuclear radiations and Calorimetry Online in Reactor' is under progress between CEA and Aix-Marseille University in the framework of a joint laboratory LIMMEX2. This program aims to improve knowledge of the neutron and photon fluxes in the RJH core, with one hand, an innovative instrumentation performing mapping of experimental locations, and on the other hand by coupling neutron flux, photon flux and thermal measurements. Neutron flux expected in the JHR core is about 10 15 n.cm -2 .s -1 and nuclear heating up to 20 W.g -1 for a nominal power of 100 MWth. One of the challenges is to identify sensors able to measure such fluxes in JHR experimental conditions and to determine how to analyse the signals delivered by these sensors with the most appropriate methods. The thesis is part of this ambitious program and aims to study the potential and the interest of the combination of radiation measurements in the prospect of a better assessment of the levels of neutron flux, gamma radiation and nuclear heating in the JHR experimental locations. The first step of IN-CORE program is to develop and operate an instrumented device called CARMEN-1 adapted to the mapping of the OSIRIS reactor, then to develop a second version called CARMEN-2 dedicated to experiments in the JHR core, especially for its start-up. This experiment was the opportunity to test all the radiation sensors which could meet the needs of JHR, including recently developed sensors. Reference neutron measurements are performed by activation

  5. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    International Nuclear Information System (INIS)

    2014-01-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  6. Characterization and testing of materials for nuclear reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2007-03-01

    Nuclear techniques in general and neutrons based methods in particular have played and will continue to play an important role in research in materials science and technology. Today the world is looking at nuclear fission and nuclear fusion as the main sources of energy supply for the future. Research reactors have played a key role in the development of nuclear technology. A materials development programme will thus play a major role in the design and development of new nuclear power plants, for the extension of the life of operating reactors as well as for fusion reactors. Against this background, the IAEA had organized a Technical Meeting on Development, Characterization and Testing of Materials - With Special Reference to the Energy Sector under the activity on specific applications of research reactors. The meeting was held in Vienna, May 29- June 2, 2006. There was also participation by experts in techniques, complementary to neutrons. The participants for the technical meeting were experts in the utilization of nuclear techniques namely the high flux and medium flux research reactors, fusion research and positron annihilation. They presented the design, development and utilization of the facilities at their respective centres for materials characterization with main focus on materials for nuclear energy, both fission and fusion. In core irradiation of materials, development of instrument for residual stress measurement in large and / or irradiated specimen, neutron radiography for inspection of irradiated fuel, work on oxide dispersion strengthened (ODS) steels and SiC composites, relevant to future power systems were cited as application of nuclear techniques in fission reactors. The use of neutron scattering for helium bubbles in steel, application of positron annihilation to study helium bubbles in Cu, Ti-stabilized stainless steel and voidswelling studies etc. show that these techniques have an important role in the development of materials for energy

  7. Fusion Reactor Safety Research program. Annual report, FY-80

    International Nuclear Information System (INIS)

    Crocker, J.G.; Cohen, S.

    1981-06-01

    The report is in three sections. Outside contracts includes a report of newly-started study at the General Atomic Company to consider safety implications of low-activation materials, portions of two papers from ongoing work at PNL and ANL, reports of the lithium spill work at HEDL, the LITFIRE code development at MIT, and risk assessment at MIT, all of which are an expansion of FY-79 outside contracts. EG and G Activities includes adaptations of four papers of ongoing work in transient code development, tritium system risk assessment, heat transfer and fluid flow analysis, and fusion safety data base. Program Plan Development includes the Executive Summary of the Plan, which was completed in FY-80, and is accompanied by a list of publications and a brief outline of proposed FY-81 activities to be based on the Program Plan

  8. Program on MOX fuel utilization in light water reactors

    International Nuclear Information System (INIS)

    Kenda, Hirofumi

    2000-01-01

    MOX fuel utilization program by the Japanese electric power companies was released in February, 1997. Principal philosophy for MOX fuel design is that MOX fuel shall be compatible with Uranium fuel and behavior of core loaded with MOX fuel shall be similar to that of conventional core. MOX fuel is designed so that geometry and nuclear capability of MOX fuel are equivalent to Uranium fuel. (author)

  9. High-temperature gas-cooled reactor (HTGR): long term program plan

    International Nuclear Information System (INIS)

    1980-01-01

    The FY 1980 effort was to investigate four technology options identified by program participants as potentially viable candidates for near-term demonstration: the Gas Turbine system (HTGR-GT), reflecting its perceived compatibility with the dry-cooling market, two systems addressing the process heat market, the Reforming (HTGR-R) and Steam Cycle (HTGR-SC) systems, and a more developmental reactor system, The Nuclear Heat Source Demonstration Reactor (NHSDR), which was to serve as a basis for both the HTGR-GT and HTGR-R systems as well as the further potential for developing advanced applications such as steam-coal gasification and water splitting

  10. SAFSIM: A computer program for engineering simulations of space reactor system performance

    International Nuclear Information System (INIS)

    Dobranich, D.

    1992-01-01

    SAFSIM (System Analysis Flow SIMulator) is a FORTRAN computer program that provides engineering simulations of user-specified flow networks at the system level. It includes fluid mechanics, heat transfer, and reactor dynamics capabilities. SAFSIM provides sufficient versatility to allow the simulation of almost any flow system, from a backyard sprinkler system to a clustered nuclear reactor propulsion system. In addition to versatility, speed and robustness are primary goals of SAFSIM. The current capabilities of SAFSIM are summarized, and some illustrative example results are presented

  11. Reactor Network Synthesis Using Coupled Genetic Algorithm with the Quasi-linear Programming Method

    OpenAIRE

    Soltani, H.; Shafiei, S.; Edraki, J.

    2016-01-01

    This research is an attempt to develop a new procedure for the synthesis of reactor networks (RNs) using a genetic algorithm (GA) coupled with the quasi-linear programming (LP) method. The GA is used to produce structural configuration, whereas continuous variables are handled using a quasi-LP formulation for finding the best objective function. Quasi-LP consists of LP together with a search loop to find the best reactor conversions (xi), as well as split and recycle ratios (yi). Quasi-LP rep...

  12. Design of particle bed reactors for the space nuclear thermal propulsion program

    International Nuclear Information System (INIS)

    Ludewig, H.; Powell, J.R.; Todosow, M.; Maise, G.; Barletta, R.; Schweitzer, D.G.

    1996-01-01

    This paper describes the design for the Particle Bed Reactor (PBR) that was considered for the Space Nuclear Thermal Propulsion (SNTP) Program. The methods of analysis and their validation are outlined first. Monte Carlo methods were used for the physics analysis, several new algorithms were developed for the fluid dynamics, heat transfer and transient analysis; and commercial codes were used for the stress analysis. We carried out a critical experiment, prototypic of the PBR to validate the reactor physics; blowdown experiments with beds of prototypic dimensions were undertaken to validate the power-extraction capabilities from particle beds. In addition, materials and mechanical design concepts for the fuel elements were experimentally validated. (author)

  13. The Canadian development program for conditioning CANDU reactor wastes for disposal

    International Nuclear Information System (INIS)

    Charlesworth, D.H.; Bourns, W.T.; Buckley, L.P.

    1978-07-01

    Currently, radioactive wastes arising from the operation of Canadian nuclear reactors are placed in interim storage in concrete containment structures except for gaseous and liquid wastes containing small amounts of radioactivity which are dispersed. With the objective of replacing storage by permanent disposal, a program is underway to develop and demonstrate an integrated process for converting all reactor wastes to a stable, leach-resistant form which will immobilize the radionuclides in the waste repository. The major tool for this development is a Waste Treatment Centre, now being constructed at Chalk River Nuclear Laboratories, which will combine reverse-osmosis, incineration, evaporation and bituminizing processes. (author)

  14. DOE/NE University Program in robotics for advanced reactors

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Gonzalez, R.C.; Trivedi, M.M.; Wehe, D.K.

    1990-05-01

    The US Department of Energy has provided support to four universities and the Oak Ridge National Laboratory in order to pursue research leading to the development and deployment of advanced robotic systems capable of performing tasks that are hazardous to humans, that generate significant occupational radiation exposure, and/or whose execution times can be reduced if performed by an automated system. The goal is to develop a generation of advanced robotic systems capable of economically performing surveillance, maintenance, and repair tasks in nuclear facilities and other hazardous environments. The approach to achieving the program objective is a transition from teleoperation to the capability of autonomous operation within three successive generations of robotic systems. The strategy adopted in order to achieve the program goals in an efficient and timely manner consists in utilizing, and advancing where required, state-of-the-art robotics technology through close interaction between the universities and the manufacturers and operators of nuclear power plants. There is a potentially broad range of applications for the robotic systems developed in the course of this project. Therefore, it is expected that efforts to obtain additional support from other agencies, e.g., DOD and NASA, will be successful. Areas of cooperation with other nations (e.g., Japan, France, Germany) are being explored. This Program features a unique teaming arrangement among the Universities of Florida, Michigan, Tennessee, Texas, and the Oak Ridge National Laboratory, and their industrial partners, Odetics, Gulf State Utilities, Florida Power and Light Company, Remotec, and Telerobotics International

  15. Liquid Metal Fast Breeder Reactor program. Volume IV. Environmental statement

    International Nuclear Information System (INIS)

    1974-12-01

    A broad overview is presented of the many implications of LMFBR program implementation, up to and encompassing a fully developed LMFBR power plant economy, including the secondary impacts, the unavoidable adverse environmental impacts, cumulative environmental impacts, and cost-benefit analyses, and alternative energy strategies. Under the heading of secondary impacts, the national implications of the availability of electricity from LMFBRs, and the specific economic impacts of the LMFBR program are examined. The currently feasible alternatives and potential future alternatives for mitigating adverse environmental impacts of the LMFBR fuel cycle are described. The problems of safeguarding special nuclear material from potential diversion to unauthorized purposes are analyzed. The cumulative environmental effects of LMFBR operation to the Year 2020, the decommissioning of LMFBRs and fuel cycle facilities upon the completion of their useful life, the irreversible and irretrievable commitments of resources that will accompany implementation of an LMFBR economy, and an analysis of the costs and benefits of implementing the LMFBR Program are included. (U.S.)

  16. University of Florida--US Department of Energy 1994-1995 reactor sharing program

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1996-06-01

    The grant support of $24,250 (1994-95?) was well used by the University of Florida as host institution to support various educational institutions in the use of UFTR Reactor. All users and uses were screened to assure the usage was for educational institutions eligible for participation in the Reactor Sharing Program; where research activities were involved, care was taken to assure the research was not funded by grants for contract funding from outside sources. Over 12 years, the program has been a key catalyst for renewing utilization of UFTR both by external users around the State of Florida and the Southeast and by various faculty members within the University of Florida. Tables provide basic information about the 1994-95 program and utilization of UFTR

  17. 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

  18. LEU fuel fabrication program for the RECH-1 reactor. Status report

    International Nuclear Information System (INIS)

    Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

    2000-01-01

    In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

  19. 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

  20. 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.

  1. Performance Demonstration Program Plan for the WIPP Experimental-Waste Characterization Program

    International Nuclear Information System (INIS)

    1991-02-01

    The Performance Demonstration Program is designed to ensure that compliance with the Quality Assurance Objective, identified in the Quality Assurance Program Plan for the WIPP Experimental-Waste Characterization Program (QAPP), is achieved. This Program Plan is intended for use by the WPO to assess the laboratory support provided for the characterization of WIPP TRU waste by the storage/generator sites. Phase 0 of the Performance Demonstration Program encompasses the analysis of headspace gas samples for inorganic and organic components. The WPO will ensure the implementation of this plan by designating an independent organization to coordinate and provide technical oversight for the program (Program Coordinator). Initial program support, regarding the technical oversight and coordination functions, shall be provided by the USEPA-ORP. This plan identifies the criteria that will be used for the evaluation of laboratory performance, the responsibilities of the Program Coordinator, and the responsibilities of the participating laboratories. 5 tabs

  2. 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

  3. The current state of the Russian reduced enrichment research reactors program

    Energy Technology Data Exchange (ETDEWEB)

    Aden, V.G.; Kartashov, E.F.; Lukichev, V.A. [and others

    1997-08-01

    During the last year after the 16-th International Conference on Reducing Fuel Enrichment in Research Reactors held in October, 1993 in Oarai, Japan, the conclusive stage of the Program on reducing fuel enrichment (to 20% in U-235) in research reactors was finally made up in Russia. The Program was started late in 70th and the first stage of the Program was completed by 1986 which allowed to reduce fuel enrichment from 80-90% to 36%. The completion of the Program current stage, which is counted for 5-6 years, will exclude the use of the fuel enriched by more than 20% from RF to other countries such as: Poland, Czeck Republick, Hungary, Roumania, Bulgaria, Libya, Viet-Nam, North Korea, Egypt, Latvia, Ukraine, Uzbekistan and Kazakhstan. In 1994 the Program, approved by RF Minatom authorities, has received the status of an inter-branch program since it was admitted by the RF Ministry for Science and Technical Policy. The Head of RF Minatom central administrative division N.I.Ermakov was nominated as the Head of the Russian Program, V.G.Aden, RDIPE Deputy Director, was nominated as the scientific leader. The Program was submitted to the Commission for Scientific, Technical and Economical Cooperation between USA and Russia headed by Vice-President A. Gore and Prime Minister V. Chemomyrdin and was given support also.

  4. 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

  5. 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.

  6. Linear programming optimization of nuclear energy strategy with sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Lee, Je Whan; Jeong, Yong Hoon; Chang, Yoon Il; Chang, Soon Heung

    2011-01-01

    Nuclear power has become an essential part of electricity generation to meet the continuous growth of electricity demand. A Sodium-cooled Fast Reactor (SFR) was developed to extend uranium resource utilization under a growing nuclear energy scenario while concomitantly providing a nuclear waste management solution. Key questions in this scenario are when to introduce SFRs and how many reactors should be introduced. In this study, a methodology using Linear Programming is employed in order to quantify an optimized growth pattern of a nuclear energy system comprising light water reactors and SFRs. The optimization involves tradeoffs between SFR capital cost premiums and the total system U3O8 price premiums. Optimum nuclear growth patterns for several scenarios are presented, as well as sensitivity analyses of important input parameters

  7. Planning and implementation of Istanbul Technical University TRIGA research reactor program

    International Nuclear Information System (INIS)

    Aybers, N.; Yavuz, H.; Bayulken, A.

    1982-01-01

    The Istanbul Technical University TRIGA Research Reactor at the Institute for Nuclear Energy, which went critical on March 11, 1979 is basically a pulsing type TRIGA Mark - II reactor. Completion of the ITU-TRR contributed to broaden the role of the Institute for Nuclear Energy of the Technical University in Istanbul in the nuclear field by providing for the first time adequate on-campus experimental facilities for nuclear engineering studies to ITU students. The research program which is currently under planning at ITU-NEE encompasses: a) Neutron activation analysis studies by techniques and applications to chemistry, mining, materials research, archaeological and biomedical studies; b) applications of Radioisotopes; c) Radiography with reactor neutron beams; d) Radiation Pulsing

  8. 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)

  9. Status of the DOE's foreign research reactor spent nuclear fuel acceptance program

    International Nuclear Information System (INIS)

    Chacey, K.; Saris, E.C.

    1997-01-01

    In May 1996, the U.S. Department of Energy (DOE), in consultation with the U.S. Department of State (DOS), adopted a policy to accept and manage in the United States ∼20 tonnes of spent nuclear fuel from research reactors in up to 41 countries. This spent fuel is being accepted under the nuclear weapons non-proliferation policy concerning foreign research reactor spent nuclear fuel. Only spent fuel containing uranium enriched in the United States is covered under this policy. Implementing this policy is a top priority of the DOE. The purpose of this paper is to provide the current status of the foreign research reactor acceptance program, including achievements to date and future challenges

  10. 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)

  11. Environmental characterization of two potential locations at Hanford for a new production reactor

    Energy Technology Data Exchange (ETDEWEB)

    Watson, E.C.; Becker, C.D.; Fitzner, R.E.; Gano, K.A.; Imhoff, K.L.; McCallum, R.F.; Myers, D.A.; Page, T.L.; Price, K.R.; Ramsdell, J.V.; Rice D.G.; Schreiber D.L.; Skumatz L.A.; Sommer D.J.; Tawil J.J.; Wallace R.W.; Watson D.G.

    1984-09-01

    This report describes various environmental aspects of two areas on the Hanford Site that are potential locations for a New Production Reactor (NPR). The area known as the Skagit Hanford Site is considered the primary or reference site. The second area, termed the Firehouse Site, is considered the alternate site. The report encompasses an environmental characterization of these two potential NPR locations. Eight subject areas are covered: geography and demography; ecology; meteorology; hydrology; geology; cultural resources assessment; economic and social effects of station construction and operation; and environmental monitoring. 80 refs., 68 figs., 109 tabs.

  12. Fracture toughness master curve characterization of Linde 1092 weld metal for Beaver valley 1 reactor

    International Nuclear Information System (INIS)

    Lee, Bong Sang; Yang, Won Jon; Hong, Jun Hwa

    2000-12-01

    This report summarizes the test results obtained from the Korean contribution to the integrity assessment of low toughness Beaver Valley reactor vessel by characterizing the fracture toughness of Linde 1092 (No. 305414) weld metal. 10 PCVN specimens and 10 1T-CT specimens were tested in accordance with the ASTM E 1921-97 standard, 'Standard test method for determination of reference temperature, T o , for ferritic steels in the transition range'. This results can also be useful for assessment of Linde 80 low toughness welds of Kori-1

  13. Fracture toughness master curve characterization of Linde 1092 weld metal for Beaver valley 1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Bong Sang; Yang, Won Jon; Hong, Jun Hwa

    2000-12-01

    This report summarizes the test results obtained from the Korean contribution to the integrity assessment of low toughness Beaver Valley reactor vessel by characterizing the fracture toughness of Linde 1092 (No. 305414) weld metal. 10 PCVN specimens and 10 1T-CT specimens were tested in accordance with the ASTM E 1921-97 standard, 'Standard test method for determination of reference temperature, T{sub o}, for ferritic steels in the transition range'. This results can also be useful for assessment of Linde 80 low toughness welds of Kori-1.

  14. Decontamination flange film characterization for a boiling water reactor under hydrogen water chemistry

    International Nuclear Information System (INIS)

    Baston, V.F.; Garbauskas, M.F.; Bozeman, J.

    1996-01-01

    Stainless steel artifacts removed from a boiling water reactor class 4 plant that operated under hydrogen water chemistry and experienced a difficult decontamination were submitted for oxide film characterization. The results reported for the corrosion film composition and structure are consistent with existing theoretical concepts for stainless steel corrosion, spinel structure site preferences (octahedral or tetrahedral) for transition metal ions, and potential-pH diagrams. The observed zinc effects on film stability and lower cobalt incorporation are also consistent with these theoretical concepts

  15. 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

  16. Nuclear R and D program in Indonesia and selection of future research reactor to support it

    International Nuclear Information System (INIS)

    Baiquni, A.; Subki, I.

    1981-01-01

    The nuclear R and D program selection decision is described as a phased program, each phase having its specific objective. The elements of each phase are identified and related with the objective, from which the activities of each element are also broadly outlined. To support the nuclear R and D program and to realize the objectives in each phase, the research facilities are also developed. A new nuclear development center housing a multipurpose reactor (MPR) and various laboratories are also described. The choice of the MPR and its criteria are also described briefly

  17. INDRA: a program system for calculating the neutronics and photonics characteristics of a fusion reactor blanket

    International Nuclear Information System (INIS)

    Perry, R.T.; Gorenflo, H.; Daenner, W.

    1976-01-01

    INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU) [de

  18. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Parma Edward J.

    2016-01-01

    Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

  19. New Production Reactors Program: Report to the Congress by the Secretary of Energy

    International Nuclear Information System (INIS)

    1990-01-01

    In August 1988, DOE issued a report to Congress announcing the Department's proposal for assuring future tritium production capacity through the building of new facilities, specifically, a preferred duality strategy that calls for the design and construction of a new, up-to-date heavy water reactor (HWR) at the Savannah River Site in South Carolina; and, at the same time, to proceed with the design and construction of a modular high-temperature gas-cooled reactor (MHTGR) plant to be located at the Idaho National Engineering Laboratory in Idaho. As a contingency, the Department is developing a light water reactor (LWR) tritium target for possible use in the partially completed WNP-1 reactor located on the Hanford Site, near Richland, Washington. The Secretary created the Office of New Production Reactors (NP) to accomplish the following mission: To provide new production reactor capacity on an urgent schedule for an assured supply of nuclear materials, primarily tritium, to maintain the Nation's nuclear deterrent capability. The organizational philosophy of the Program calls for strong central control with increased safety, environmental, and quality assurance oversight. In addition to separate Engineering Technology Offices for HWR, MHTGR, and LWR that will maintain technical control throughout the life of the Program, the NP Director has created oversight Offices for Safety and Quality and for Environment. An NP Project Management Office will be established at each of the three project sites with the Project Officer reporting directly to the Director of NP. Safety, Quality, and Environmental Field Representatives will be located on-site at the Project Management Offices to serve as a ''resident'' for the Headquarters' Office to whom they will report

  20. The neutron utilization and promotion program of TRR-II research reactor project in Taiwan

    International Nuclear Information System (INIS)

    Gone, J.K.; Huang, Y.H.

    2001-01-01

    The objective of the Taiwan research reactor system improvement and utilization promotion project is to reconstruct the old Taiwan research reactor (TRR), which was operated by the Institute of Nuclear Energy Research (INER) between 1973 and 1988, into a multi-purpose medium flux research reactor (TRR-II). The project started in 1998, and the new reactor is scheduled to have its first critical in June of 2006. The estimated maximum unperturbed thermal neutron flux (E 14 n/cm 2 sec, and it is about one order of magnitude higher than other operating research reactors in Taiwan. The new reactor will equip with secondary neutron sources to provide neutrons with different energies, which will be an essential tool for advanced material researches in Taiwan. One of the major tasks of TRR-II project is to promote domestic utilization of neutrons generated at TRR-II. The traditional uses of neutrons in fuel/material research, trace element analysis, and isotope production has been carried out at INER for many years. On the other hand, it is obvious that promotions of neutron spectrometric technique will be a major challenge for the project team. The limited neutron flux from operating research reactors had discouraged domestic users in developing neutron spectrometric technique for many years, and only few researchers in Taiwan are experienced in using spectrometers. It is important for the project team to encourage domestic researchers to use neutron spectrometers provided by TRR-II as a tool for their future researches in various fields. This paper describes the current status of TRR-II neutron utilization and promotion program. The current status and future plans for important issues such as staff recruiting, personnel training, international collaboration, and promotion strategy will be described. (orig.)

  1. Updated Generation IV Reactors Integrated Materials Technology Program Plan, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Corwin, William R [ORNL; Burchell, Timothy D [ORNL; Halsey, William [Lawrence Livermore National Laboratory (LLNL); Hayner, George [Idaho National Laboratory (INL); Katoh, Yutai [ORNL; Klett, James William [ORNL; McGreevy, Timothy E [ORNL; Nanstad, Randy K [ORNL; Ren, Weiju [ORNL; Snead, Lance Lewis [ORNL; Stoller, Roger E [ORNL; Wilson, Dane F [ORNL

    2005-12-01

    The Department of Energy's (DOE's) Generation IV Nuclear Energy Systems Program will address the research and development (R&D) necessary to support next-generation nuclear energy systems. Such R&D will be guided by the technology roadmap developed for the Generation IV International Forum (GIF) over two years with the participation of over 100 experts from the GIF countries. The roadmap evaluated over 100 future systems proposed by researchers around the world. The scope of the R&D described in the roadmap covers the six most promising Generation IV systems. The effort ended in December 2002 with the issue of the final Generation IV Technology Roadmap [1.1]. The six most promising systems identified for next generation nuclear energy are described within the 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. 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 the structural materials needed to ensure their safe and reliable operation. Accordingly, DOE has identified materials as one of the focus areas for Gen IV technology development.

  2. Fusion reactor materials program plan. Section III. Plasma material interaction

    International Nuclear Information System (INIS)

    1978-07-01

    A discussion of materials-related problems and an analysis of such problems is given for each major topical area. The strategy that will be used to solve the materials problems is described. As part of this program strategy, a series of major milestones is identified that extends over the next 20 years. Detailed task descriptions for the next five years leading to the achievement of the major milestones are given. Each task is described on a separate page (or task sheet) which includes the task number, task title, objective, scope, and the major milestones addressed by the task. Secondary milestones within a given task or subtask are defined, together with a priority assignment and an estimate of man-years to accomplish the work. Each Plan is organized along major topics which parallel the Subtask organization of the Task Group responsible for the Plan

  3. Generic aging management programs for license renewal of BWR reactor coolant systems components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  4. Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  5. 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)

  6. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Directory of Open Access Journals (Sweden)

    Gendron T.

    2011-04-01

    Full Text Available In order for New Brunswick Power Nuclear (NBPN to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS the development of an aging management plan (AMP was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  7. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Science.gov (United States)

    Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

    2011-04-01

    In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  8. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Malkoske, G.R.

    2002-01-01

    Historically, the production of molybdenum-99 in the NRU research reactors at Chalk River, Canada has been extracted from reactor targets employing highly enriched uranium (HEU). A reliable supply of HEU metal from the United States used in the manufacture of targets for the NRU research reactor has been a key factor to enable MDS Nordion to develop a secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets provides predictable, consistent yields for our high-volume molybdenum production process. Each link of the isotope supply chain, from isotope production to ultimate use by the physician, has been established using this proven and established method of HEU target irradiation and processing to extract molybdenum-99. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the construction of two MAPLE reactors and a New Processing Facility. The design of the MAPLE facilities was based on an established process developed by Atomic Energy of Canada Ltd. (AECL) - extraction of isotopes from HEU target material. However, in concert with the global trend to utilize low enriched uranium (LEU) in research reactors, MDS Nordion has launched a three phase LEU Target Development and Conversion Program for the MAPLE facilities. Phase 1, the Initial Feasibility Study, which identified the technical issues to convert the MAPLE reactor targets from HEU to LEU for large scale commercial production was reported on at the RERTR- 2000 conference. The second phase of the LEU Target Development and Conversion Program was developed with extensive consultation and involvement of experts knowledgeable in target development, process system design, enriched uranium conversion chemistry and commercial scale reactor operations and molybdenum production. This paper will provide an overview of the Phase 2 Conversion Development Program, report on progress to date, and further

  9. Participation in the US Department of Energy Reactor Sharing Program. Annual report, September 30, 1993--September 29, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  10. Development of a risk-based inservice inspection program for a liquid metal reactor

    International Nuclear Information System (INIS)

    King, R.W.; Buschman, H.W.

    1996-01-01

    The emerging application of risk-based assessment technology to the operation and maintenance of nuclear power plants holds considerable promise for improving efficiency and reducing operating costs. EBR-II is liquid-metal-cooled fast reactor which operated for thirty years before shutting down in September 1994 due to program termination. Prior to the shutdown of EBR-II, an in-service inspection (ISI) program was developed that exploited certain advantages of the liquid-metal reactor design, e.g., demonstrated passive response to plant upset events, low pressure primary coolant and compatibility of the coolant and reactor materials. Many of the systems cannot be inspected due to inaccessibility of the components. However, application of a risk-based approach provided the basis for reducing or eliminating inspections in some areas that would otherwise be required. Development and implementation of the risk-based ISI program was interrupted by the DOE-mandated shutdown of EBR-II, so the potential benefits of this approach in terms of reduced O and M costs have yet to be realized. Through the development of this program, however it is clear that there is potential for substantial cost-savings while improving the risk-profile of the facility through this approach

  11. Advancement in reactor coolant chemistry management programs and related technology development in Taiwan

    International Nuclear Information System (INIS)

    Huang, C.S.; Lin, Chien C.

    2000-01-01

    Taiwan Power Company (TPC) has three nuclear power plants in operation with a total capacity of 51 GWe, contributing about 30% of electricity generation in Taiwan. The first two plants, Chinshan (CSNPP) and Kuosheng (KSNPP), are boiling water reactor plants, and the third one, Maanshan (MASNPP), is a pressurized water reactor plant. Each plant has two identical reactors. As many nuclear power plant operators worldwide, TPC is committed to operate the plants efficiently, economically, and safely. TPC has developed and implemented several chemistry improvement programs in recent years to improve the coolant chemistry in order to ( l ) protect structure materials from corrosion, (2) reduce radiation exposures to workers and (3) reduce radwaste production and radiation release to the environment. This paper describes TPC's experience in some water chemistry management, radwaste reduction and radiation exposure control programs. Future programs under planning, including implementation of hydrogen water chemistry (HWC) in BWRs, installation of condensate pre-filters, and development of on-line water chemistry monitoring system, are also be briefly discussed. In addition, some material related research and development programs will also be presented. (author)

  12. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, October 1, 1982-March 3, 1983

    International Nuclear Information System (INIS)

    1983-06-01

    This report provides descriptions and results of the technical effort during the first half of FY 83 on the Gas-Cooled Thermal Reactor Program. The work on Integration and Management (WBS 01) includes the preparation of the Advanced Systems Concept Evaluation Plan and the Advanced Systems Technology Development Plan in addition to the program management activities. The Market Definition (WBS 03) efforts considered the application of the Modular Reactor System with reforming (MRS-R) to the production of methanol and ammonia and the refining of petroleum. Within the Plant Technology (WBS 13) task there were activities to develop anlytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. In addition to the work on the advanced HTGR for process heat users, new activities were initiated in support of the HTGR-SC/C Lead plant Protect (WBS 30 and 31). The Plant Simulation task (WBS 31) was initiated to develop a computer code for simulation of plant operation and for plant transient systems analysis. The efforts on the advanced HTGR systems was performed under the Modular Systems task (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors

  13. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  14. Light Water Reactor Sustainability Program: Digital Architecture Project Plan

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    There are many technologies available to the nuclear power industry to improve efficiency in plant work activities. These range from new control room technologies to those for mobile field workers. They can make a positive impact on a wide range of performance objectives – increase in productivity, human error reduction, validation of results, accurate transfer of data, and elimination of repetitive tasks. It is expected that the industry will more and more turn to these technologies to achieve these operational efficiencies to lower costs. At the same time, this will help utilities manage a looming staffing problem as the inevitable retirement wave of the more seasoned workers affects both staffing levels and knowledge retention. A barrier to this wide-scale implementation of new technologies for operational efficiency is the lack of a comprehensive digital architecture that can support the real-time information exchanges needed to achieve the desired operational efficiencies. This project will define an advanced digital architecture that will accommodate the entire range of system, process, and plant worker activity to enable the highest degree of integration, thereby creating maximum efficiency and productivity. This pilot project will consider a range of open standards that are suitable for the various data and communication requirements of a seamless digital environment. It will map these standards into an overall architecture to support the II&C developments of this research program.

  15. Dosimetry program for characterization of the FMIT facility

    International Nuclear Information System (INIS)

    Gold, R.; McElroy, W.N.; Fuller, J.L.; Sheen, E.M.; Dierckx, R.

    1979-01-01

    The environmental characterization program for the Fusion Materials Irradiaton Test (FMIT) facility is presented. Requirements for the development and testing of Magnetic Fusion Energy (MFE) materials together with the complexity of the FMIT (d,Li) generated radiation field warrant a multifaceted dosimetric approach. Specific passive, active and calculational dosimetry efforts comprising this multifaceted approach are described. Special emphasis is given to those dosimetry capabilities uniquely required to characterize FMIT

  16. Adaptive sampling program support for expedited site characterization

    International Nuclear Information System (INIS)

    Johnson, R.

    1993-01-01

    Expedited site characterizations offer substantial savings in time and money when assessing hazardous waste sites. Key to some of these savings is the ability to adapt a sampling program to the ''real-time'' data generated by an expedited site characterization. This paper presents a two-prong approach to supporting adaptive sampling programs: a specialized object-oriented database/geographical information system for data fusion, management and display; and combined Bayesian/geostatistical methods for contamination extent estimation and sample location selection

  17. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

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

  18. Neutron study of fast neutron reactor systems by exponential experiments on Harmonie - Graphite program HUG-PHUG - Oxide program PHRIXOS - Uranium program UK

    International Nuclear Information System (INIS)

    Desprets, Alain.

    1977-12-01

    Exponential experiments allow to obtain the fundamental characteristics of a lattice (material buckling, reaction rate ratios) more economically than critical experiments. This report describes the experimental techniques and the methods of analysis used for this type of experiments. The results obtained with three programs performed with the source reactor HARMONIE are given: graphite-lattices program (3 U-fueled and 3 Pu-fueled lattices); oxide-fuel program (4 PuO 2 -UO 2 lattices); pure uranium program (one lattice). Some of these lattices were also studied in critical experiments. The coherence of the results obtained by the two types of experiments is established [fr

  19. Deriving a site characterization program from applicable regulations

    International Nuclear Information System (INIS)

    Voegele, M.D.; Younker, J.L.; Alexander, D.H.

    1988-01-01

    The process of deriving a site characterization program from the applicable regulations was approached by the DOE through the use of two basic organizing principles. One organizing principle is a hierarchical structure of questions about regulatory criteria related to the acquisition of site data. This set of questions is called an issues hierarchy, and it provides a topical organizing framework for developing a site characterization program. The second basic organizing principle used by the DOE and its contractors to develop a site characterization program is called performance allocation. For each issue in the issues hierarchy, a resolution strategy is developed. These strategies involve the identification of elements of the disposal system that are relevant to isolation and containment of waste or to radiological safety. It is then possible to identify performance measures and information needed from the site characterization program. This information, coupled with information about confidence in existing data and the confidence required in the data to be obtained, allows the development of testing strategies for field programs

  20. Hydrodynamic characterization and evaluation of an open channel reactor for the degradation of paracetamol

    International Nuclear Information System (INIS)

    Abreu Zamora, Maria A.; Gonzalez Lopez, Dagoberto E.; Robaina Leon, Yalaina; Dominguez Catasus, Judith; Borroto Portela, Jorge I.; Jauregui Haza, Ulises J.

    2015-01-01

    The conventional wastewater treatment plants do not guarantee the degradation of Persistent Organic Pollutants (POPs). Advanced oxidation processes, like photodegradation that use artificial ultraviolet and solar radiation, are proposed as an alternative for the treatment of contaminated water with POPs. In the present work, the hydrodynamic characterization and evaluation of an open channel reactor for the degradation of paracetamol are presented. The hydrodynamic characterization was performed through the analysis of the residence time distribution using a radioisotope 99m Tc. This process was done in two steps. First, the open channel reactor was evaluated in continuous mode operation. To study the influence of the fluid volume in the reactor and the diameter of the flow distributor's orifices on the flow pattern, an experimental 3 2 design with two replicas in the center was used. The dependent variables were the number of perfectly mixed tanks (J), the mean residence time of the model (τ) and the experimental mean residence time (Trm). The model of perfectly mixed tanks in series exchanging with stagnant zones was assumed as the best model. In a second moment, the mixing time of the system operating in close loop mode was determined. Finally, the degradation of paracetamol in aqueous dissolution trough photolysis, photolysis intensified with H 2 O 2 , photo-Fenton with artificial ultraviolet radiation and photo-Fenton with solar radiation was evaluated. The results show that the photo-Fenton processes employing artificial ultraviolet and solar radiation warranty the total degradation of the pharmaceutical after 15 minutes of reaction. (Author)