Science.gov

Sample records for reactor safety division

  1. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  2. Water Reactor Safety Research Division quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Romano, A.J.

    1980-06-01

    The Water Reactor Safety Research Programs Quarterly Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evaluation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  3. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  4. Health, Safety, and Environment Division

    SciTech Connect

    Wade, C

    1992-01-01

    The primary responsibility of the Health, Safety, and Environmental (HSE) Division at the Los Alamos National Laboratory is to provide comprehensive occupational health and safety programs, waste processing, and environmental protection. These activities are designed to protect the worker, the public, and the environment. Meeting these responsibilities requires expertise in many disciplines, including radiation protection, industrial hygiene, safety, occupational medicine, environmental science and engineering, analytical chemistry, epidemiology, and waste management. New and challenging health, safety, and environmental problems occasionally arise from the diverse research and development work of the Laboratory, and research programs in HSE Division often stem from these applied needs. These programs continue but are also extended, as needed, to study specific problems for the Department of Energy. The results of these programs help develop better practices in occupational health and safety, radiation protection, and environmental science.

  5. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  6. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  7. High-temperature gas-cooled reactor safety studies for the Division of Reactor Safety Research. Quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Harrington, R.M.

    1980-08-01

    Work continued on development of the ORTAP, ORECA, and BLAST codes; and verification studies were continued on the ORECA, CORTAP, and BLAST codes. An improved steam turbine plant model (ORTURB) for use in ORTAP was developed and checked. Predictions from BLAST, CORTAP, and ORECA were compared with various transient test data from the Fort St. Vrain reactor.

  8. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  9. Reactor safety assessment system

    SciTech Connect

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

  10. Nuclear reactor safety device

    DOEpatents

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  11. Perspectives on reactor safety

    SciTech Connect

    Haskin, F.E.; Camp, A.L.

    1994-03-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  12. N Reactor operational safety summary

    SciTech Connect

    Franz, G.R.; Quapp, W.J.; Ogden, D.M.

    1988-08-01

    This report is a safety summary of the N Reactor. Beginning with its conceptual design in the mid-1950`s, and throughout its 23 years of operation, continuous efforts have been made to ensure safe N Reactor operation and protection of the public health and safety. The N Reactor Updated Safety Analysis Report, completed in 1978(UNC1978), and its subsequent amendments document the safety bases of N Reactor. Following the April 1986 Chernobyl accident in the Soviet Union, a major effort to confirm N Reactor safety and further increase its safety margin was initiated. This effort, called the Safety Enhancement Program, reassessed the N Reactor using the latest accepted analysis techniques and commercial light-water reactor guidelines, where applicable. 122 refs., 38 figs., 10 tabs.

  13. Nuclear reactor safety device

    DOEpatents

    Hutter, E.

    1983-08-15

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  14. Reactor operation safety information document

    SciTech Connect

    Not Available

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  15. Physics of reactor safety. Quarterly report, October-December 1980. Volume IV

    SciTech Connect

    Not Available

    1981-02-01

    The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

  16. Physics of reactor safety. Quarterly report, July-September 1980. Volume III

    SciTech Connect

    Not Available

    1980-11-01

    This Quarterly progress report summarizes work done during the months of July-September 1980 in Argonne National Laboratory's Applied Physics and Components Technology Divisions for the Division of Reactor Safety Research of the US Nuclear Regulatory Commission. The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions. An executive summary is provided including a statement of the findings and recommendations of the report.

  17. Nuclear Reactor Safety: a current awareness bulletin

    SciTech Connect

    Cunningham, D.C.

    1985-01-15

    Nuclear Reactor Safety announces on a semimonthly basis the current worldwide information available on all safety-related aspects of fission reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures.

  18. Thermal reactor safety

    SciTech Connect

    Not Available

    1980-06-01

    Information is presented concerning new trends in licensing; seismic considerations and system structural behavior; TMI-2 risk assessment and thermal hydraulics; statistical assessment of potential accidents and verification of computational methods; issues with respect to improved safety; human factors in nuclear power plant operation; diagnostics and activities in support of recovery; LOCA transient analysis; unresolved safety issues and other safety considerations; and fission product transport.

  19. Thermal Reactor Safety

    SciTech Connect

    Not Available

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  20. Automatic safety rod for reactors

    DOEpatents

    Germer, John H.

    1988-01-01

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  1. [Reactor safety and human failure].

    PubMed

    Smidt, D

    1979-12-01

    Reactor safety is given by the reliable solution of 3 tasks: on-time shutdown, continuous decay-heat removal, safe containment. After describing the general strategy of their solution even under upset conditions the most important engineered safeguards of pressurized water reactors are summarized. The important problem of human failure is discussed in some more detail. For the example Harrisburg some difficulties, but also some technical countermeasures are illustrated. PMID:537639

  2. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  3. Health, Safety, and Environment Division annual report 1989

    SciTech Connect

    Wade, C.

    1992-01-01

    The primary responsibility of the Health, Safety, and Environment (HSE) Division at the Los Alamos National Laboratory is to provide comprehensive occupational health and safety programs, waste processing, and environmental protection. These activities are designed to protect the worker, the public, and the environment. Meeting the responsibilities involves many disciplines, including radiation protection, industrial hygiene, safety, occupational medicine, environmental science and engineering, analytical chemistry, epidemiology, and waste management. New and challenging health, safety, and environmental problems occasionally arise from the diverse research and development work of the Laboratory, and research programs in the HSE Division often stem from these applied needs. These programs continue but are also extended, as needed, to study specific problems for the Department of Energy. The result of these programs is to help develop better practices in occupational health and safety, radiation protection, and environmental sciences.

  4. Health, Safety, and Environment Division annual report, 1988

    SciTech Connect

    Rosenthal, M.A.

    1989-10-01

    The primary responsibility of the Health, Safety, and Environment (HSE) Division at the Los Alamos National Laboratory is to provide comprehensive occupational health and safety programs, waste processing, and environmental protection. These activities are designed to protect the worker, the public, and the environment. Many disciplines are required to meet the responsibilities, including radiation protection, industrial hygiene, safety, occupational medicine, environmental science, epidemiology, and waste management. New and challenging health and safety problems occasionally arise from the diverse research and development work of the Laboratory. Research programs in HSE Division often stem from these applied needs. These programs continue but are also extended, as needed, to study specific problems for the Department of Energy and to help develop better occupational health and safety practices. 52 refs.

  5. Health, Safety, and Environment Division: Annual progress report 1987

    SciTech Connect

    Rosenthal, M.A.

    1988-04-01

    The primary responsibility of the Health, Safety, and Environment (HSE) Division at the Los Alamos National Laboratory is to provide comprehensive occupational health and safety programs, waste processing, and environment protection. These activities are designed to protect the worker, the public, and the environment. Many disciplines are required to meet the responsibilities, including radiation protection, industrial hygiene, safety, occupational medicine, environmental science, epidemiology, and waste management. New and challenging health and safety problems arise occasionally from the diverse research and development work of the Laboratory. Research programs in HSE Division often stem from these applied needs. These programs continue but are also extended, as needed to study specific problems for the Department of Energy and to help develop better occupational health and safety practices.

  6. Review of light water reactor safety

    SciTech Connect

    Cheng, H.S.

    1980-12-01

    A review of the present status of light water reactor (LWR) safety is presented. The review starts with a brief discussion of the outstanding accident scenarios concerning LWRs. Where possible the areas of present technological uncertainties are stressed. To provide a better perspective of reactor safety, it then reviews the probabilistic assessment of the outstanding LWR accidents considered in the Reactor Safety Study (WASH-1400) and discusses the potential impact of the present technological uncertainties on WASH-1400.

  7. Monitoring circuit for reactor safety systems

    DOEpatents

    Keefe, Donald J.

    1976-01-01

    The ratio between the output signals of a pair of reactor safety channels is monitored. When ratio falls outside of a predetermined range, it indicates that one or more of the safety channels has malfunctioned.

  8. Safety control circuit for a neutronic reactor

    DOEpatents

    Ellsworth, Howard C.

    2004-04-27

    A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

  9. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

  10. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

  11. MP-Division health and safety reference handbook. [Contains glossary

    SciTech Connect

    Putnam, T.M.

    1987-09-01

    This report presents the objectives, organization, policies, and essential rules and procedures that have been adopted by MP Division and that form the basis of the Health and Safety Program of the Clinton P. Anderson Meson Physics Facility (LAMPF). The facility includes the beam-delivery systems for the Los Alamos Neutron Scattering Center and the Weapons Neutron Research Facility (LANSCE/WNR). The program is designed not only to assure the health and safety of all personnel, including users, in their work at LAMPF, and of MP-Division staff in their work on the LANSCE/WNR beam lines, but also to protect the facility (buildings and equipment) and the environment. 33 refs., 18 figs., 2 tabs.

  12. Computational chemistry in Argonne`s Reactor Analysis Division

    SciTech Connect

    Gelbard, E.; Agrawal, R.; Fanning, T.

    1997-08-01

    Roughly 3 years ago work on Argonne`s Integral Fast Reactor ({open_quotes}IFR{close_quotes}) was terminated and at that time, ANL funding was redirected to a number of alternative programs. One such alternative was waste management and, since disposal of spent fuel from ANL`s EBR-II reactor presents some special problems, this seemed an appropriate area for ANL work. Methods for the treatment and disposal of spent fuel (particularly from EBR-II but also from other sources) are now under very active investigation at ANL. The very large waste form development program is mainly experimental at this point, but within the Reactor Analysis ({open_quotes}RA{close_quotes}) Division a small computational chemistry program is underway, designed to supplement the experimental program. One of the most popular proposals for the treatment of much of our high-level wastes is vitrification. As noted below, this approach has serious drawbacks for EBR-II spent fuel. ANL has proposed, instead, that spent fuel first be pretreated by a special metallurgical process which produces, as waste, chloride salts of the various fission products; these salts would then be adsorbed in zeolite A, which is subsequently bonded with glass to produce a waste form suitable for disposal. So far it has been the main mission of RA`s computational chemistry program to study the process by which leaching occurs when the glass-bonded zeolite waste form is exposed to water. It is the purpose of this paper to describe RA`s computational chemistry program, to discuss the computational techniques involved in such a program, and in general to familiarize the M. and C. Division with a computational area which is probably unfamiliar to most of its member. 11 refs., 2 figs.

  13. NEW EMPLOYEES ON THE JOB - DONALD E HEGBERG OF THE NUCLEAR REACTOR DIVISION DISCUSSES NUCLEAR ROCKET

    NASA Technical Reports Server (NTRS)

    1963-01-01

    NEW EMPLOYEES ON THE JOB - DONALD E HEGBERG OF THE NUCLEAR REACTOR DIVISION DISCUSSES NUCLEAR ROCKET FUEL ELEMENT EXPERIMENT WITH CHARLES L YOUNGER - THE DISCUSSION IS PREPATORY TO CONDUCTING THE EXPERIMENT AT THE PLUM BROOK STATION REACTOR FACILITY

  14. Reactor Safety Research Programs Quarterly Report April- June 1981

    SciTech Connect

    Edler, S. K.

    1981-09-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory {INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  15. Some views on nuclear reactor safety

    SciTech Connect

    Tanguy, P.Y.

    1995-04-01

    This document is the text of a speech given by Pierre Y. Tanguy (Electricite de France) at the 22nd Water Reactor Safety Meeting held in Bethesda, MD in 1994. He describes the EDF nuclear program in broad terms and proceeds to discuss operational safety results with EDF plants. The speaker also outlines actions to enhance safety planned for the future, and he briefly mentions French cooperation with the Chinese on the Daya Bay project.

  16. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect

    R. Wigeland; J. Cahalan

    2009-09-01

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

  17. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect

    Hosseini, Seyed Mohammad Hadi; Nematollahi, Mohammad Reza; Sepanloo, Kamran

    2004-07-01

    Probabilistic Safety Assessment (PSA) application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this paper the application of the Probabilistic Safety Assessment to the Tehran Research Reactor (TRR) is presented. The level 1 PSA application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using SAPHIRE software. This Study shows that the obtained core damage frequency for Tehran Research Reactor (8.368 E-6 per year) well meets the IAEA criterion for existing nuclear power plants (1E-4). But safety improvement suggestions are offered to decrease the most probable accidents. (authors)

  18. Evolution of space reactor safety

    SciTech Connect

    Wetch, J.R. )

    1993-01-10

    This paper summarizes some of the major safety issues and proposed and developed solutions to real and perceived problems. Safety of the public, operating personnel, facilities and the environment has always been a driving concern of the public, government agencies, system designers and users of nuclear power and propulsion systems.

  19. Reactor Safety Research Programs Quarterly Report April -June 1980

    SciTech Connect

    Edler, S. K.

    1980-11-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  20. Reactor Safety Research Programs Quarterly Report October - December 1980

    SciTech Connect

    Edler, S K

    1981-04-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  1. Reactor Safety Research Programs Quarterly Report January - March 1980

    SciTech Connect

    Hagen, C. M

    1980-10-01

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  2. Automatic safety rod for reactors. [LMFBR

    DOEpatents

    Germer, J.H.

    1982-03-23

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  3. Perspectives on reactor safety. Revision 1

    SciTech Connect

    Haskin, F.E.; Camp, A.L.; Hodge, S.A.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  4. Austrian contributions to reactor safety research

    NASA Astrophysics Data System (ADS)

    Sdouz, Gert

    1992-03-01

    An overview of Austrian contributions to reactor safety research is given. Starting with licensing for the power plant Zwenkendorf and participation in the safety projects PBF (Power Burst Facility) and LOFT (Loss Of Fluid Test), the work shifted later to participation in the study of international standard problems of the OECD and the IAEA. Since 1988 the center of activity is the calculation of the source term behavior of accident sequences, especially for reactors of Russian types. As an illustration the results of a source term calculation for a TMLB' sequence are presented.

  5. Reactor safety for the Space Exploration Initiative

    NASA Technical Reports Server (NTRS)

    Dix, Terry E.

    1991-01-01

    A task force was created by the National Aeronautics and Space Administration to conduct a 90-day study to support efforts to determine requirements to meet the goals of the Space Exploration Initiative. The task force identified the need for a nuclear reactor to provide the electrical power required as the outpost power demands on the moon and Mars evolve into hundreds of kilowatts. A preliminary hazards analysis has been performed to examine safety aspects of nuclear reactor power systems for representative missions to the moon and Mars. Mission profiles were defined for reference lunar and Martian flights. Potential alternatives to each mission phase were also defined. Accident scenarios were qualitatively defined for the mission phases. The safety issues decay heat removal, reactor control, disposal, criticality, end-of-mission shutdown, radiation exposure, the Martian environment, high speed impact on the surfaces of the moon or Mars, and return flyby trajectories were identified.

  6. Trends in fusion reactor safety research

    NASA Astrophysics Data System (ADS)

    Herring, J. S.; Holland, D. F.; Piet, S. J.

    Fusion has the potential to be an attractive energy source. From the safety and environmental perspective, fusion must avoid concerns about catastrophic accidents and unsolvable waste disposal. In addition, fusion must achieve an acceptable level of risk from operational accidents that result in public exposure and economic loss. Finally, fusion reactors must control routine radioactive effluent, particularly tritium. Major progress in achieving this potential rests on development of low-activation materials or alternative fuels. The safety and performance of various material choices and fuels for commercial fusion reactors can be investigated relatively inexpensively through reactor design studies. These studies bring together experts in a wide range of backgrounds and force the group to either agree on a reactor design or identify areas for further study. Fusion reactors will be complex, with distributed radioactive inventories. The next generation of experiments will be critical in demonstrating that acceptable levels of safe operation can be achieved. These machines will use materials which are available today and for which a large database exists (e.g., for 316 stainless steel). Researchers have developed a good understanding of the risks associated with operation of these devices. Specifically, consequences from coolant system failures, loss of vacuum events, tritium releases, and liquid metal reactions have been studied. Recent studies go beyond next step designs and investigate commercial reactor concerns including tritium release and liquid metal reactions.

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

    SciTech Connect

    Not Available

    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.

  8. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety.

  9. Small Liquid Metal Cooled Reactor Safety Study

    SciTech Connect

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  10. 78 FR 48468 - Delphi Corporation, Electronics and Safety Division, Including On-Site Leased Workers From...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-08

    ... determination was published in the Federal Register on June 10, 2013 (Volume 78 FR Pages 34672-34674). At the... Employment and Training Administration Delphi Corporation, Electronics and Safety Division, Including On-Site... to workers of Delphi Corporation, Electronics and Safety Division, including on-site leased...

  11. Natural circulation reactor design safety analysis

    NASA Astrophysics Data System (ADS)

    Zheng, Dong

    2001-07-01

    This thesis study covers both global performance and local phenomena analyses focusing on natural circulation reactor design safety. Four important topics are included: the global SBWR design safety assessment, important local phenomena investigation, steady and transient natural circulation process study, and two-phase instability analysis. The conceptual design of the SBWR-200 is introduced in this thesis and the global performance of a natural circulation reactor is then assessed using PUMA integral test data and RELAP5 simulations. A safety assessment methodology is developed to evaluate the PUMA integral test data extrapolation and code scalability. The RELAP5 code simulation capability in low-pressure low-flow conditions is also validated. The study shows that the code is capable of predicting the global accident scenario in natural circulation reactors with reasonable accuracy, while failing to reproduce some safety related local phenomena. The natural circulation process is investigated in detail using PUMA separate effect natural circulation tests. The natural circulation flow rate and heat transfer rate have been modeled analytically and numerically. The work indicates that two-phase natural circulation has enough capability to remove decay power. However, the flow instability observed in two-phase natural circulation cases seriously challenges the feasibility of natural circulation reactor design. The instability is classified as a type of density wave instability induced by flashing. A detailed stability study is performed focusing on flashing induced instability under natural circulation condition. Various flashing phenomena have been studied and a mechanistic flashing model has been proposed and improved using a relaxation method. The developed relaxation flashing model can be applied to general two-phase non-equilibrium phenomena.

  12. Fusion reactor breeder material safety compatibility studies

    SciTech Connect

    Jeppson, D.W.; Cohen, S.; Muhlestein, L.D.

    1983-09-01

    Tritium breeder material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Breeder material safety compatibility studies are being conducted to identify and characterize breeder-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate the following. 1. Ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/, and LiTiO/sub 3/) at postulated blanket operating temperatures are chemically compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ reactions with water generate heat, aerosol, and hydrogen. 2. Lithium oxide and 17Li-83Pb alloy react mildly with water requiring special precautions to control hydrogen release. 3. Liquid lithium reacts substantially, while 17Li83Pb alloy reacts mildly with concrete to produce hydrogen. 4. Liquid lithium-air reactions may present some major safety concerns. Additional scoping tests are needed, but the ternary oxides, lithium oxide, and 17Li-83Pb have definite safety advantages over liquid lithium and Li/sub 7/Pb/sub 2/. The ternary oxides present minimal safetyrelated problems when used with water as coolant, air or concrete; but they do require neutron multipliers, which may have safety compatibility concerns with surrounding materials. The combined favorable neutronics and minor safety compatibility concerns of lithium oxide and 17Li-83Pb make them prime candidates as breeder materials. Current safety efforts are directed toward assessing the compatibility of lithium oxide and the lithium-lead alloy with coolants and other materials.

  13. Passive Safety Features for Small Modular Reactors

    SciTech Connect

    Ingersoll, Daniel T

    2010-01-01

    The rapid growth in the size and complexity of commercial nuclear power plants in the 1970s spawned an interest in smaller, simpler designs that are inherently or intrinsically safe through the use of passive design features. Several designs were developed, but none were ever built, although some of their passive safety features were incorporated into large commercial plant designs that are being planned or built today. In recent years, several reactor vendors are actively redeveloping small modular reactor (SMR) designs with even greater use of passive features. Several designs incorporate the ultimate in passive safety they completely eliminate specific accident initiators from the design. Other design features help to reduce the likelihood of an accident or help to mitigate the accident s consequences, should one occur. While some passive safety features are common to most SMR designs, irrespective of the coolant technology, other features are specific to water, gas, or liquid-metal cooled SMR designs. The extensive use of passive safety features in SMRs promise to make these plants highly robust, protecting both the general public and the owner/investor. Once demonstrated, these plants should allow nuclear power to be used confidently for a broader range of customers and applications than will be possible with large plants alone.

  14. Space reactor safety, 1985--1995 lessons learned

    SciTech Connect

    Marshall, A.C.

    1995-12-31

    Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

  15. The next generation of power reactors - safety characteristics

    SciTech Connect

    Modro, S.M.

    1995-01-01

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

  16. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-08-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  17. Issues affecting advanced passive light-water reactor safety analysis

    SciTech Connect

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented.

  18. Job task and functional analysis of the Division of Reactor Projects, office of Nuclear Reactor Regulation. Final report

    SciTech Connect

    Morzinski, J.A.; Gilmore, W.; Hahn, H.A.

    1998-07-10

    A job task and functional analysis was recently completed for the positions that make up the regional Divisions of Reactor Projects. Among the conclusions of that analysis was a recommendation to clarify roles and responsibilities among site, regional, and headquarters personnel. As that analysis did not cover headquarters personnel, a similar analysis was undertaken of three headquarters positions within the Division of Reactor Projects: Licensing Assistants, Project Managers, and Project Directors. The goals of this analysis were to systematically evaluate the tasks performed by these headquarters personnel to determine job training requirements, to account for variations due to division/regional assignment or differences in several experience categories, and to determine how, and by which positions, certain functions are best performed. The results of this analysis include recommendations for training and for job design. Data to support this analysis was collected by a survey instrument and through several sets of focus group meetings with representatives from each position.

  19. Safety program considerations for space nuclear reactor systems

    SciTech Connect

    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.

  20. Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

    SciTech Connect

    D. Kokkinos

    2005-04-28

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

  1. Nuclear safety criteria and specifications for space nuclear reactors

    SciTech Connect

    Not Available

    1982-08-01

    The purpose of this document is to define safety criteria which must be met to implement US safety policy for space fission reactors. These criteria provide the bases for decisions on the acceptability of specific mission and reactor design proposals. (JDH)

  2. Nuclear Reactor Safety--The APS Submits its Report

    ERIC Educational Resources Information Center

    Physics Today, 1975

    1975-01-01

    Presents the summary section of the American Physical Society (APS) report on the safety features of the light-water reactor, reviews the design, construction, and operation of a reactor and outlines the primary engineered safety features. Summarizes the major recommendations of the study group. (GS)

  3. Savannah River Site K-Reactor Probabilistic Safety Assessment

    SciTech Connect

    Brandyberry, M.D.; Bailey, R.T.; Baker, W.H.; Kearnaghan, D.P.; O`Kula, K.R.; Wittman, R.S.; Woody, N.D.; Amos, C.N.; Weingardt, J.J.

    1992-12-01

    This report gives the results of a Savannah River Site (SRS) K-Reactor Probabilistic Safety Assessment (PSA). Measures of adverse consequences to health and safety resulting from representations of severe accidents in SRS reactors are presented. In addition, the report gives a summary of the methods employed to represent these accidents and to assess the resultant consequences. The report is issued to provide useful information to the U. S. Department of Energy (DOE) on the risk of operation of SRS reactors, for insights into severe accident phenomena that contribute to this risk, and in support of improved bases for other DOE programs in Heavy Water Reactor safety.

  4. Safety evaluation report related to the renewal of the facility license for the research reactor at the Dow Chemical Company

    SciTech Connect

    Not Available

    1989-04-01

    This safety evaluation report for the application filed by the Dow Chemical Company for renewal of facility Operating License R-108 to continue to operate its research reactor at an increased operating power level has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the grounds of the Michigan Division of the Dow Chemical Company in Midland, Michigan. The staff concludes that the Dow Chemical Company can continue to operate its reactor without endangering the health and safety of the public.

  5. Technology Development, Evaluation, and Application (TDEA) FY 2001 Progress Report Environment, Safety, and Health (ESH) Division

    SciTech Connect

    L.G. Hoffman; K. Alvar; T. Buhl; E. Foltyn; W. Hansen; B. Erdal; P. Fresquez; D. Lee; B. Reinert

    2002-05-01

    This progress report presents the results of 11 projects funded ($500K) in FY01 by the Technology Development, Evaluation, and Application (TDEA) Committee of the Environment, Safety, and Health Division (ESH). Five projects fit into the Health Physics discipline, 5 projects are environmental science and one is industrial hygiene/safety. As a result of their TDEA-funded projects, investigators have published sixteen papers in professional journals, proceedings, or Los Alamos reports and presented their work at professional meetings. Supplement funds and in-kind contributions, such as staff time, instrument use, and workspace, were also provided to TDEA-funded projects by organizations external to ESH Divisions.

  6. Reactor Safety Research: Semiannual report, July-December 1986

    SciTech Connect

    Not Available

    1987-11-01

    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 the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and 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 tehcnology base supporting licensing decisions.

  7. VVER Reactor Safety in Eastern Europe and Former Soviet Union

    NASA Astrophysics Data System (ADS)

    Papadopoulou, Demetra

    2012-02-01

    VVER Soviet-designed reactors that operate in Eastern Europe and former Soviet republics have heightened international concern for years due to major safety deficiencies. The governments of countries with VVER reactors have invested millions of dollars toward improving the safety of their nuclear power plants. Most of these reactors will continue to operate for the foreseeable future since they provide urgently-needed electrical power. Given this situation, this paper assesses the radiological consequences of a major nuclear accident in Eastern Europe. The paper also chronicles the efforts launched by the international nuclear community to improve the safety of the reactors and notes the progress made so far through extensive collaborative efforts in Armenia, Bulgaria, the Czech Republic, Hungary, Kazakhstan, Lithuania, Russia, Slovakia, and Ukraine to reduce the risks of nuclear accidents. Western scientific and technical staff collaborated with these countries to improve the safety of their reactor operations by strengthening the ability of the regulator to perform its oversight function, installing safety equipment and technologies, investing time in safety training, and working diligently to establish an enduring safety culture. Still, continued safety improvement efforts are necessary to ensure safe operating practices and achieve timely phase-out of older plants.

  8. Study Gives Good Odds on Nuclear Reactor Safety

    ERIC Educational Resources Information Center

    Russell, Cristine

    1974-01-01

    Summarized is data from a recent study on nuclear reactor safety completed by Norman C. Rasmussen and others. Non-nuclear events are about 10,000 times more likely to produce large accidents than nuclear plants. (RH)

  9. Health and Safety Research Division progress report for the period October 1, 1991--March 31, 1993

    SciTech Connect

    Berven, B.A.

    1993-09-01

    This is a progress report from the Health and Safety Research Division of Oak Ridge National Laboratory. Information is presented in the following sections: Assessment Technology, Biological and Radiation Physics, Chemical Physics, Biomedical and Environmental Information Analysis, Risk Analysis, Center for Risk Management, Associate Laboratories for Excellence in Radiation Technology (ALERT), and Contributions to National and Lead Laboratory Programs and Assignments--Environmental Restoration.

  10. Industrial Safety. MAS-123. Waste Isolation Division (WID). Management and Supervisor Training (MAST) Program.

    ERIC Educational Resources Information Center

    Westinghouse Electric Corp., Carlsbad, NM.

    This learning module, which is part of a management and supervisor training program for managers and supervisors employed at the Department of Energy's Waste Isolation Division, is designed to prepare trainees to promote and monitor the industrial safety program at their plant. The following topics are covered in the module's individual sections:…

  11. Nuclear-safety criteria and specifications for space nuclear reactors

    SciTech Connect

    Not Available

    1982-08-01

    The policy of the United States for all US nuclear power sources in space is to ensure that the probability of release of radioactive material and the amounts released are such that an undue risk is not presented, considering the benefits of the mission. The objective of this document is to provide safety criteria which a mission/reactor designer can use to help ensure that the design is acceptable from a radiological safety standpoint. These criteria encompass mission design, reactor design, and radiological impact limitation requirements for safety, and the documentation required. They do not address terrestrial operations, occupational safety or system reliability except where the systems are important for radiological safety. Specific safety specifications based on these criteria shall also be generated and made part of contractual requirements.

  12. Technology Development, Evaluation, and Application (TDEA) FY 1999 Progress Report, Environment, Safety, and Health (ESH) Division

    SciTech Connect

    Larry G. Hoffman

    2000-12-01

    This progress report presents the results of 10 projects funded ($500K) in FY99 by the Technology Development, Evaluation, and Application (TDEA) Committee of the Environment, Safety, and Health Division. Five are new projects for this year; seven projects have been completed in their third and final TDEA-funded year. As a result of their TDEA-funded projects, investigators have published thirty-four papers in professional journals, proceedings, or Los Alamos reports and presented their work at professional meetings. Supplemental funds and in-kind contributions, such as staff time, instrument use, and work space, were also provided to TDEA-funded projects by organizations external to ESH Division.

  13. Generation III reactors safety requirements and the design solutions

    NASA Astrophysics Data System (ADS)

    Felten, P.

    2009-03-01

    Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called "generation III" safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

  14. Generation III reactors safety requirements and the design solutions

    SciTech Connect

    Felten, P.

    2009-03-31

    Nuclear energy's public acceptance, and hence its development, depends on its safety. As a reactor designer, we will first briefly remind the basic safety principles of nuclear reactors' design. We will then show how the industry, and in particular Areva with its EPR, made design evolution in the wake of the Three Miles Island accident in 1979. In particular, for this new generation of reactors, severe accidents are taken into account beyond the standard design basis accidents. Today, Areva's EPR meets all so-called 'generation III' safety requirements and was licensed by several nuclear safety authorities in the world. Many innovative solutions are integrated in the EPR, some of which will be introduced here.

  15. Safety status of space radioisotope and reactor power sources

    NASA Technical Reports Server (NTRS)

    Bennett, Gary L.

    1990-01-01

    The current overall safety criterion for both radioisotope and reactor power sources is containment or immobilization in the case of a reentry accident. In addition, reactors are designed to remain subcritical under conditions of land impact or water immersion. A very extensive safety test and analysis program was completed on the radioisotope thermoelectric generators (RTGs) in use on the Galileo spacecraft and planned for use on the Ulysses spacecraft. The results of this work show that the RTGs will pose little or no risk for any credible accident. The SP-100 space nuclear reactor program has begun addressing its safety criteria, and the design is planned to be such as to ensure meeting the various safety criteria. Preliminary mission risk analyses on SP-100 show the expected value population dose from postulated accidents on the reference mission to be very small. It is concluded that the current US nuclear power sources are the safest flown.

  16. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect

    Lawrence, J.D.

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  17. Preliminary safety evaluation of the advanced burner test reactor.

    SciTech Connect

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  18. Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

    SciTech Connect

    T. A. Tomberlin; S. B. Grover

    2004-11-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

  19. Simulation of the modified K reactor supplementary safety system

    SciTech Connect

    Paik, I.K.; Canas, L.R. ); Peterson, P.F. )

    1991-01-01

    The supplementary safety system (SSS) of the K reactor provides a second line of defense to shut down the reactor if the safety and control rods fail to scram. The SSS was originally designed to inject a neutron poison solution (ink) into the reactor tank via spargers. Recently, concerns arose that the ink inventory might run out before the ink front returned to the moderator during a loss-of-ac-power transient in which the coolant pumps coast down. Thus, a new system has been added to inject additional ink through the pump suctions so that ink will arrive in the core before depletion of the sparger ink. The MODFLOW code was developed to calculate the moderator flow distribution in Savannah River site (SRS) reactors, including the effects of inertia and stratification from buoyancy forces.

  20. Safety philosophy of gas turbine high temperature reactor (GTHTR300)

    SciTech Connect

    Shoji Katanishi; Kazuhiko Kunitomi; Shusaku Shiozawa

    2002-07-01

    Japan Atomic Energy Research Institute (JAERI) has undertaken the study of an original design concept of gas turbine high temperature reactor, the GTHTR300. The general concept of this study is development of a greatly simplified design that leads to substantially reduced technical and cost requirements. Newly proposed design features enable the GTHTR300 to be an efficient and economically competitive reactor in 2010's. Also, the GTHTR300 fully takes advantage of its inherent safety characteristics. The safety philosophy of the GTHTR300 is developed based on the HTTR (High Temperature Engineering Test Reactor) of JAERI which is the first HTGR in Japan. Major features of the newly proposed safety philosophy for the GTHTR300 are described in this article. (authors)

  1. Hydrogen and water reactor safety: proceedings

    SciTech Connect

    Not Available

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  2. METHOD AND APPARATUS FOR REACTOR SAFETY CONTROL

    DOEpatents

    Huston, N.E.

    1961-06-01

    A self-contained nuclear reactor fuse controlled device tron absorbing material, normally in a compact form but which can be expanded into an extended form presenting a large surface for neutron absorption when triggered by an increase in neutron flux, is described.

  3. New reactor technology: safety improvements in nuclear power systems.

    PubMed

    Corradini, M L

    2007-11-01

    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems. PMID:18049233

  4. Methods and strategies for future reactor safety goals

    NASA Astrophysics Data System (ADS)

    Arndt, Steven Andrew

    There have been significant discussions over the past few years by the United States Nuclear Regulatory Commission (NRC), the Advisory Committee on Reactor Safeguards (ACRS), and others as to the adequacy of the NRC safety goals for use with the next generation of nuclear power reactors to be built in the United States. The NRC, in its safety goals policy statement, has provided general qualitative safety goals and basic quantitative health objectives (QHOs) for nuclear reactors in the United States. Risk metrics such as core damage frequency (CDF) and large early release frequency (LERF) have been used as surrogates for the QHOs. In its review of the new plant licensing policy the ACRS has looked at the safety goals, as has the NRC. A number of issues have been raised including what the Commission had in mind when it drafted the safety goals and QHOs, how risk from multiple reactors at a site should be combined for evaluation, how the combination of a new and old reactor at the same site should be evaluated, what the criteria for evaluating new reactors should be, and whether new reactors should be required to be safer than current generation reactors. As part of the development and application of the NRC safety goal policy statement the Commissioners laid out the expectations for the safety of a nuclear power plant but did not address the risk associated with current multi-unit sites, potential modular reactor sites, and hybrid sites that could contain current generation reactors, new passive reactors, and/or modular reactors. The NRC safety goals and the QHOs refer to a "nuclear power plant," but do not discuss whether a "plant" refers to only a single unit or all of the units on a site. There has been much discussion on this issue recently due to the development of modular reactors. Additionally, the risk of multiple reactor accidents on the same site has been largely ignored in the probabilistic risk assessments (PRAs) done to date, and in most risk

  5. 76 FR 63676 - Final Division of Safety Systems Interim Staff Guidance DSS-ISG-2010-01: Staff Guidance Regarding...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-13

    ... COMMISSION Final Division of Safety Systems Interim Staff Guidance DSS-ISG- 2010-01: Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools AGENCY: Nuclear Regulatory Commission... Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools.'' This DSS-ISG provides...

  6. NUSAR: N Reactor Updated Safety Analysis Report, Amendment 21

    SciTech Connect

    Smith, G L

    1989-12-01

    The enclosed pages are Amendment 21 of the N Reactor Updated Safety Analysis Report (NUSAR). NUSAR, formerly UNI-M-90, was revised by 18 amendments that were issued by UNC Nuclear Industries, the contractor previously responsible for N Reactor operations. As of June 1987, Westinghouse Hanford Company (WHC) acquired the operations and engineering contract for N Reactor and other facilities at Hanford. The document number for NUSAR then became WHC-SP-0297. The first revision was issued by WHC as Amendment 19, prepared originally by UNC. Summaries of each of the amendments are included in NUSAR Section 1.1.

  7. Transactions of the nineteenth water reactor safety information meeting

    SciTech Connect

    Weiss, A.J.

    1991-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.

  8. Reactor Safety Research Programs Quarterly Report July - September 1981

    SciTech Connect

    Edler, S. K.

    1982-01-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  9. Reactor Safety Research Programs Quarterly Report October - December 1981

    SciTech Connect

    Edler, S. K.

    1982-03-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  10. An Overview of the Safety Case for Small Modular Reactors

    SciTech Connect

    Ingersoll, Daniel T

    2011-01-01

    Several small modular reactor (SMR) designs emerged in the late 1970s and early 1980s in response to lessons learned from the many technical and operational challenges of the large Generation II light-water reactors. After the accident at the Three Mile Island plant in 1979, an ensuing reactor redesign effort spawned the term inherently safe designs, which later evolved into passively safe terminology. Several new designs were engineered to be deliberately small in order to fully exploit the benefits of passive safety. Today, new SMR designs are emerging with a similar philosophy of offering highly robust and resilient designs with increased safety margins. Additionally, because these contemporary designs are being developed subsequent to the September 11, 2001, terrorist attack, they incorporate a number of intrinsic design features to further strengthen their safety and security. Several SMR designs are being developed in the United States spanning the full spectrum of reactor technologies, including water-, gas-, and liquid-metal-cooled ones. Despite a number of design differences, most of these designs share a common set of design principles to enhance plant safety and robustness, such as eliminating plant design vulnerabilities where possible, reducing accident probabilities, and mitigating accident consequences. An important consequence of the added resilience provided by these design approaches is that the individual reactor units and the entire plant should be able to survive a broader range of extreme conditions. This will enable them to not only ensure the safety of the general public but also help protect the investment of the owner and continued availability of the power-generating asset. Examples of typical SMR design features and their implications for improved plant safety are given for specific SMR designs being developed in the United States.

  11. Health and Safety Research Division progress report, October 1, 1988--March 31, 1990

    SciTech Connect

    Not Available

    1990-09-01

    The Health and Safety Research Division (HASRD) of the Oak Ridge National Laboratory (ORNL) continues to maintain an outstanding program of basic and applied research displaying a high level of creativity and achievement as documented by awards, publications, professional service, and successful completion of variety of projects. Our focus is on human health and the scientific basis for measurement and assessment of health-related impacts of energy technologies. It is our custom to publish a division progress report every 18 months that summarizes our programmatic progress and other measures of achievement over the reporting period. Since it is not feasible to summarize in detail all of our work over the period covered by this report (October 1, 1988, to March 30, 1990), we intend this document to point the way to the expensive open literature that documents our findings. During the reporting period the Division continued to maintain strong programs in its traditional areas of R D, but also achieved noteworthy progress in other areas. Much of the Division's work on site characterization, development of new field instruments, compilation of data bases, and methodology development fits into this initiative. Other new work in tunneling microscopy in support of DOE's Human Genome Program and the comprehensive R D work related to surface-enhanced Raman spectroscopy have attained new and exciting results. These examples of our progress and numerous other activities are highlighted in this report.

  12. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  13. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  14. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  15. The temperature effect and safety of the Topaz reactor

    SciTech Connect

    Artyukhov, G.Y.; Zelentsov, S.N.; Ionkin, V.I.; Kudryavtsev, V.P.; Makarenkov, Y.D.; Marin, S.N.; Pupko, V.Y.; Raskach, F.P. )

    1991-01-01

    This paper presents the results of a design calculation to compute the 3-dimensional geometry of the TOPAZ reactor. Emphasis was given to the calculation of the temperature effects and temperature reactivity coefficients. Performance of the control and safety systems, spectral and dynamic characteristics were also evaluated.(AIP)

  16. Nuclear safety as applied to space power reactor systems

    SciTech Connect

    Cummings, G.E.

    1987-01-01

    To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper.

  17. Performance and safety parameters for the high flux isotope reactor

    SciTech Connect

    Ilas, G.; Primm III, T.

    2012-07-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

  18. Performance and Safety Parameters for the High Flux Isotope Reactor

    SciTech Connect

    Ilas, Germina; Primm, Trent

    2012-01-01

    A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data.

  19. Technology Development, Evaluation, and Application (TDEA) FY 1998 Progress Report Environment, Safety, and Health (ESH) Division

    SciTech Connect

    Larry G. Hoffman; Kenneth Alvar; Thomas Buhl; Bruce Erdal; Philip Fresquez; Elizabeth Foltyn; Wayne Hansen; Bruce Reinert

    1999-06-01

    This progress report presents the results of 10 projects funded ($504K) in FY98 by the Technology Development, Evaluation, and Application (TDEA) Committee of the Environment, Safety, and Health Division. Nine projects are new for this year; two projects were completed in their third and final TDEA-funded year. As a result of their TDEA-funded projects, investigators have published 19 papers in professional journals, proceedings, or Los Alamos reports and presented their work at professional meetings. Supplemental funds and in-kind contributions, such as staff time, instrument use, and work space were also provided to the TDEA-funded projects by organizations external to ESH Division. Products generated from the projects funded in FY98 included a new extremity dosimeter that replaced the previously used finger-ring dosimeters, a light and easy-to-use detector to measure energy deposited by neutron interactions, and a device that will allow workers to determine the severity of a hazard.

  20. Reactor safety research programs. Quarterly report, April-June 1982

    SciTech Connect

    Edler, S.K.

    1982-11-01

    This document summarizes work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1982, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  1. Reactor safety research programs. Quarterly report, January-March 1982

    SciTech Connect

    Edler, S.K.

    1982-07-01

    This document summarizes work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1982, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities.

  2. Lessons learned from commercial reactor safety analysis

    NASA Astrophysics Data System (ADS)

    Fragola, J. R.

    1992-07-01

    As design concepts involving nuclear power are developed for space missions, prudence requires a consideration of the historical perspective provided by the commerical nuclear power generating station industry. This would allow the aerospace industry to take advantage of relevant historical experience, drawing from the best features and avoiding the pitfalls which appear to have stifled the growth of the commercial nuclear industry as a whole despite its comparatively admirable safety performance record. This paper provides some history of the development of commercial nuclear plant designs, and discusses the lessons which have been learned and how they apply to the space nuclear propulsion situation.

  3. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Gregg L. Sharp; R. T. McCracken

    2003-06-01

    The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  4. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect

    Sharp, G.L.; McCracken, R.T.

    2003-05-13

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  5. Health and Safety Research Division progress report for the period April 1, 1990--September 30, 1991

    SciTech Connect

    Kaye, S.V.

    1992-03-01

    This is a brief progress report from the Health and Safety Research Division of Oak Ridge National Laboratory. Information is presented in the following sections: Assessment Technology including Measurement Applications and Development, Pollutant Assessments, Measurement Systems Research, Dosimetry Applications Research, Metabolism and Dosimetry Research and Nuclear Medicine. Biological and Radiation Physics including Atomic, Molecular, and High Voltage Physics, Physics of Solids and Macromolecules, Liquid and Submicron Physics, Analytic Dosimetry and Surface Physics and Health Effects. Chemical Physics including Molecular Physics, Photophysics and Advanced Monitoring Development. Biomedical and Environmental Information Analysis including Human Genome and Toxicology, Chemical Hazard Evaluation and Communication, Environmental Regulations and Remediation and Information Management Technology. Risk Analysis including Hazardous Waste.

  6. Health and Safety Research Division progress report for the period April 1, 1987--September 30, 1988

    SciTech Connect

    Kaye, S.V.

    1989-03-01

    The mission of the Health and Safety Research Division (HASRD) is to provide a sound scientific basis for the measurement and assessment of human health impacts of radiological and chemical substances. Our approach to fulfilling this mission is to conduct a broad program of experimental, theoretical, and field research based on a strong foundation of fundamental physical studies that blend into well-established programs in life sciences. Topics include biomedical screening techniques, biological and chemical sensors, risk assessment, health hazards, dosimetry, nuclear medicine, environmental pollution monitoring, electron-molecule interactions, interphase physics, surface physics, data base management, environmental mutagens, carcinogens, and tetratogens.

  7. Foundational development of an advanced nuclear reactor integrated safety code.

    SciTech Connect

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  8. Space Reactor Launch Safety--An Acceptably Low Risk

    SciTech Connect

    Weitzberg, Abraham; Wright, Steven

    2008-01-21

    Results from previous space reactor and radioisotope power source risk assessments were combined to provide a scoping assessment of the possible risks from the launch of a reactor power system for use on the surface of the moon or Mars. It is assumed that future reactor power system launches would be subject to the same rigorous safety analysis and launch approval process as past nuclear payload launches. Using the same methodology that has gained approval of past launches, it was determined that the mission risk would be 0.029 person-rem worldwide which translates to 1.5*10{sup -5} latent health effects. It is seen that the only significant sources of radiological risks from a non-operating reactor are possible inadvertent criticality accidents and the consequences of such events have been shown to be extremely low. Passive means such as spectral shift poisons or high reactor core length/diameter ratios have been shown to be able to reduce or eliminate the possibility of the more credible criticality accidents, such as flooding or sand burial. This paper advances the premise that, for design purposes, future space reactor surface-power designs should primarily address the credible accidents and not the hypothetical accidents. For launch accidents and other safety assessments, a probabilistic risk assessment approach will have to be used to assess the safety impact of all types of accidents, including the hypothetical accidents. With this approach, the design of the system will not be burdened with design features that are based on hypothetical criticality accidents having negligible risk. Moreover, there is little chance of convincingly demonstrating that these design features can substantially reduce or eliminated the risk associated with hypothetical criticality accidents.

  9. Space Reactor Launch Safety-An Acceptably Low Risk

    NASA Astrophysics Data System (ADS)

    Weitzberg, Abraham; Wright, Steven

    2008-01-01

    Results from previous space reactor and radioisotope power source risk assessments were combined to provide a scoping assessment of the possible risks from the launch of a reactor power system for use on the surface of the moon or Mars. It is assumed that future reactor power system launches would be subject to the same rigorous safety analysis and launch approval process as past nuclear payload launches. Using the same methodology that has gained approval of past launches, it was determined that the mission risk would be 0.029 person-rem worldwide which translates to 1.5*10-5 latent health effects. It is seen that the only significant sources of radiological risks from a non-operating reactor are possible inadvertent criticality accidents and the consequences of such events have been shown to be extremely low. Passive means such as spectral shift poisons or high reactor core length/diameter ratios have been shown to be able to reduce or eliminate the possibility of the more credible criticality accidents, such as flooding or sand burial. This paper advances the premise that, for design purposes, future space reactor surface-power designs should primarily address the credible accidents and not the hypothetical accidents. For launch accidents and other safety assessments, a probabilistic risk assessment approach will have to be used to assess the safety impact of all types of accidents, including the hypothetical accidents. With this approach, the design of the system will not be burdened with design features that are based on hypothetical criticality accidents having negligible risk. Moreover, there is little chance of convincingly demonstrating that these design features can substantially reduce or eliminated the risk associated with hypothetical criticality accidents.

  10. PWR systems transient analysis: a reactor-safety perspective

    SciTech Connect

    Kennedy, M.F.; Abramson, P.B.; McDonald, T.A.

    1982-01-01

    In the simulation of transient events in large PWR reactor systems for reactor safety studies, the plant model is quite detailed and must include most of the plant components and control systems to adequately analyze the range of transients. The results discussed were calculated with the RELAP4/MOD6 code and reveal the need for the analysis to carefully review and understand the results to assure that they are not being adversely affected by the improper solution techniques or changes in models during the calculation.

  11. Safety significance of ATR (Advanced Test Reactor) passive safety response attributes

    SciTech Connect

    Atkinson, S.A.

    1989-01-01

    The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory was designed with some passive safety response attributes which contribute to the safety posture of the facility. The three passive safety attributes being evaluated in the paper are: (1) In-core and in-vessel natural convection cooling, (2) a passive heat sink capability of the ATR primary coolant system (PCS) for the transfer of decay power from the uninsulated piping to the confinement, and (3) gravity feed of emergency coolant makeup. The safety significance of the ATR passive safety response attributes is that the reactor can passively respond for most transients, given a reactor scram, to provide adequate decay power removal and a significant time for operator action should the normal active heat removal systems and their backup systems both fail. The ATR Interim Level 1 Probabilistic Risk Assessment (PRA) model ands results were used to evaluate the significance to ATR fuel damage frequency (or probability) of the above three passive response attributes. The results of the evaluation indicate that the first attribute is a major safety characteristic of the ATR. The second attribute has a noticeable but only minor safety significance. The third attribute has no significant influence on the ATR Level 1 PRA because of the diversity and redundancy of the ATR firewater injection system (emergency coolant system). 8 refs., 4 figs., 1 tab.

  12. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    SciTech Connect

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  13. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    SciTech Connect

    Diamond, D. J.; Baek, J. S.; Hanson, A. L.; Cheng, L-Y; Brown, N.; Cuadra, A.

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  14. Summary of HEDL Fusion Reactor Safety Support studies

    SciTech Connect

    Muhlestein, L.D.; Jeppson, D.W.; Barreca, J.R.

    1981-01-01

    The HEDL Fusion Reactor Safety Support studies are focused on characterizing blanket-coolant-material reactions for deuterium-tritium fusion reactor designs. The objective is to determine and examine potential safety and environmental issues associated with proposed blanket/coolant combinations under postulated accident conditions. The first studies considered liquid lithium as both blanket and coolant, and examined liquid lithium-material reactions. Liquid lithium reactions with oxygen, nitrogen, and various concretes have been characterized. Evaluations of lithium reaction extinguishment methods, lithium aerosol generation and collection, and the volatilization and transport of radioactive materials in connection with lithium-air reactions have been completed. Lithium compound blanket material reactions with water, a prime coolant candidate, have been characterized in terms of energy and gas release rates. Blanket materials considered were lithium aluminate, lithium oxide, lithium zirconate, lithium silicate, and lithium lead alloys (Li/sub 7/Pb/sub 2/ and Li/sub 17/Pb/sub 83/).

  15. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    SciTech Connect

    Not Available

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  16. Safety of New Generation Concepts in Reference to Present Reactors

    SciTech Connect

    Rouyer, Jean-Loup; Vitton, Francis

    2004-07-01

    Present operational nuclear reactors have reached a high level of safety and reliability. Main safety principles which have allowed to obtain excellent present performances must remain the cornerstone for future projects, with adaptations on points where operating feedback shows needs for reinforcement or simplification. These basic principles are defense in depth concept and probabilistic quantification. Criteria for which specific emphasis should be put for the future are: Inertia of processes, important parameter contributing to stability, applying to neutronics as well as coolant fluids. Barriers, whose number must not be rigidly fixed, but determined by independence and margins considerations. Hazards, risk induced by external and internal hazards which should be at least, comparable to the one induced by internal accidents. On the basis of these safety criteria, the paper analyses the potentials of several reactor concepts for the future, as compared with advanced existing reactors. The six concepts selected by the Generation IV Forum (VHTR, GFR, SFR, LFR, SCWR, MSR) are appreciated. The MHTGR project which has been evaluated by the US-NRC is also considered as a reference for gas-cooled advanced concepts. (authors)

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

    SciTech Connect

    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.

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

    SciTech Connect

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

    1992-07-01

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

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

    SciTech Connect

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

    1992-07-01

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

  20. Progress toward international agreement to improve reactor safety

    SciTech Connect

    Lieberman, J.I.; Graham, B.

    1993-05-14

    Representatives of nearly one-half of the 114 member states of the International Atomic Energy Agency (IAEA), including the United States, have participated in the development of an international nuclear safety conventions proposed multilateral treaty to improve civil nuclear power reactor safety. A preliminary draft of the convention has been developed (referred to as the draft convention for this report), but discussions are continuing, and when the final convention text will be completed and presented to IAEA member states for signature is uncertain. This report responds to the former and current Chairman`s request that we provide information on the development of the nuclear safety convention, including a discussion of (1) the draft convention`s scope and objectives, (2) how the convention will be implemented and monitored, (3) the views of selected country representatives on what provisions should be included in the draft convention, and (4) the convention`s potential benefits and limitations.

  1. Adjoint-based sensitivity analysis for reactor safety applications

    SciTech Connect

    Parks, C.V.

    1986-08-01

    The application and usefulness of an adjoint-based methodology for performing sensitivity analysis on reactor safety computer codes is investigated. The adjoint-based methodology, referred to as differential sensitivity theory (DST), provides first-order derivatives of the calculated quantities of interest (responses) with respect to the input parameters. The basic theoretical development of DST is presented along with the needed general extensions for consideration of model discontinuities and a variety of useful response definitions. A simple analytic problem is used to highlight the general DST procedures. finally, DST procedures presented in this work are applied to two highly nonlinear reactor accident analysis codes: (1) FASTGAS, a relatively small code for analysis of a loss-of-decay-heat-removal accident in a gas-cooled fast reactor, and (2) an existing code called VENUS-II which has been employed for analyzing the core disassembly phase of a hypothetical fast reactor accident. The two codes are different both in terms of complexity and in terms of the facets of DST which can be illustrated. Sensitivity results from the adjoint codes ADJGAS and VENUS-ADJ are verified with direct recalcualtions using perturbed input parameters. The effectiveness of the DST results for parameter ranking, prediction of response changes, and uncertainty analysis are illustrated. The conclusion drawn from this study is that DST is a viable, cost-effective methodology for accurate sensitivity analysis. In addition, a useful sensitivity tool for use in the fast reactor safety area has been developed in VENUS-ADJ. Future work needs to concentrate on combining the accurate first-order derivatives/results from DST with existing methods (based solely on direct recalculations) for higher-order response surfaces.

  2. Flooding Experiments and Modeling for Improved Reactor Safety

    SciTech Connect

    Solmos, M., Hogan, K.J., VIerow, K.

    2008-09-14

    Countercurrent two-phase flow and “flooding” phenomena in light water reactor systems are being investigated experimentally and analytically to improve reactor safety of current and future reactors. The aspects that will be better clarified are the effects of condensation and tube inclination on flooding in large diameter tubes. The current project aims to improve the level of understanding of flooding mechanisms and to develop an analysis model for more accurate evaluations of flooding in the pressurizer surge line of a Pressurized Water Reactor (PWR). Interest in flooding has recently increased because Countercurrent Flow Limitation (CCFL) in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA and because analysis of hypothetical severe accidents with the current flooding models in reactor safety codes shows that these models represent the largest uncertainty in analysis of steam generator tube creep rupture. During a hypothetical station blackout without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. The flooding model heavily influences the pressurizer emptying rate and the potential for surge line structural failure due to overheating and creep rupture. The air-water test results in vertical tubes are presented in this paper along with a semi-empirical correlation for the onset of flooding. The unique aspects of the study include careful experimentation on large-diameter tubes and an integrated program in which air-water testing provides benchmark knowledge and visualization data from which to conduct steam-water testing.

  3. Westinghouse independent safety review of Savannah River production reactors

    SciTech Connect

    Leggett, W.D.; McShane, W.J. ); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. . Nuclear and Advanced Technology Div.); Toto, G. . Nuclear Services Div.); Fauske, H.K. ); Call, D.W. (Westinghouse Savannah R

    1989-04-01

    Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K,L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone.

  4. Fusion-reactor blanket-material safety-compatibility studies

    SciTech Connect

    Jeppson, D.W.; Muhlestein, L.D.; Keough, R.F.; Cohen, S.

    1982-11-01

    Blanket material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Blanket material safety compatibility studies are being conducted to identify and characterize blanket-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate that : (1) ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/ and LiTiO/sub 3/) at postulated blanket operating temperatures are compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ alloy reactions with water generate heat, aerosol and hydrogen; (2) lithium oxide and Li/sub 17/Pb/sub 83/ alloy react mildly with water requiring special precautions to control hydrogen release; (3) liquid lithium reacts substantially, while Li/sub 17/Pb/sub 83/ alloy reacts mildly with concrete to produce hydrogen; and (4) liquid lithium-air reactions present some major safety concerns.

  5. Application of software to development of reactor-safety codes

    SciTech Connect

    Wilburn, N.P.; Niccoli, L.G.

    1980-09-01

    Over the past two-and-a-half decades, the application of new techniques has reduced hardware cost for digital computer systems and increased computational speed by several orders of magnitude. A corresponding cost reduction in business and scientific software development has not occurred. The same situation is seen for software developed to model the thermohydraulic behavior of nuclear systems under hypothetical accident situations. For all cases this is particularly noted when costs over the total software life cycle are considered. A solution to this dilemma for reactor safety code systems has been demonstrated by applying the software engineering techniques which have been developed over the course of the last few years in the aerospace and business communities. These techniques have been applied recently with a great deal of success in four major projects at the Hanford Engineering Development Laboratory (HEDL): 1) a rewrite of a major safety code (MELT); 2) development of a new code system (CONACS) for description of the response of LMFBR containment to hypothetical accidents, and 3) development of two new modules for reactor safety analysis.

  6. Adjoint-based sensitivity analysis for reactor-safety applications

    SciTech Connect

    Parks, C.V.

    1985-01-01

    The application and usefulness of an adjoint-based methodology for performing sensitivity analysis on reactor safety computer codes is investigated. The adjoint-based methodology, referred to as differential sensitivity theory (DST), provides first-order derivatives of the calculated quantities of interest (responses) with respect to the input parameters. The basic theoretical development of DST is presented along with the needed general extensions for consideration of model discontinuities and a variety of useful response definitions. A simple analytic problem is used to highlight the general DST procedures. Finally, DST procedures presented in this work are applied to two highly nonlinear reactor accident analysis codes: (1) FASTGAS, a relatively small code for analysis of loss-of-decay-heat-removal accident in a gas-cooled fast reactor, and (2) an existing code called VENUS-II which is typically employed for analyzing the core disassembly phase of a hypothetical fast reactor accident. The two codes are different both in terms of complexity and in terms of the facets of DST which can be illustrated. Sensitivity results from the adjoint codes ADJGAS and VENUS-ADJ are verified with direct recalculations using perturbed input parameters. The effectiveness of the DST results for parameter ranking, prediction of response changes, and uncertainty analysis are illustrated. The conclusion drawn from this study is that DST is a viable, cost-effective methodology for accurate sensitivity analysis.

  7. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    SciTech Connect

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  8. Safety aspects of Particle Bed Reactor plutonium burner system

    SciTech Connect

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

    1993-08-01

    An assessment is made of the safety aspects peculiar to using the Particle Bed Reactor (PBR) as the burner in a plutonium disposal system. It is found that a combination of the graphitic fuel, high power density possible with the PBR and engineered design features results in an attractive concept. The high power density potentially makes it possible to complete the plutonium burning without requiring reprocessing and remanufacturing fuel. This possibility removes two hazardous steps from a plutonium burning complex. Finally, two backup cooling systems depending on thermo-electric converters and heat pipes act as ultimate heat removal sinks in the event of accident scenarios which result in loss of fuel cooling.

  9. Generic safety insights for inspection of boiling water reactors

    SciTech Connect

    Higgins, J.C.; Taylor, J.H.; Fresco, A.N.; Hillman, B.M.

    1987-01-01

    As the number of operating nuclear power plants (NPPs) increases, safety inspection has increased in importance. Over the last 2 yr, probabilistic risk assessment (PRA) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only approx. 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the Brookhaven National Lab. program to develop generic boiling water reactor (BWR) PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs.

  10. Occupational health and environment research 1983: Health, Safety, and Environment Division. Progress report

    SciTech Connect

    Voelz, G.L.

    1985-05-01

    The primary responsibility of the Health, Safety, and Environment (HSE) Division at the Los Alamos National Laboratory is to provide comprehensive occupational health and safety programs, waste processing, and environmental protection. These activities are designed to protect the workers, the public, and the environment. Evaluation of respiratory protective equipment included the XM-30 and M17A1 military masks, use of MAG-1 spectacles in respirators, and eight self-contained units. The latter units were used in an evaluation of test procedures used for Bureau of Mines approval of breathing apparatuses. Analyses of air samples from field studies of a modified in situ oil shale retorting facility were performed for total cyclohexane extractables and selected polynuclear aromatic hydrocarbons. Aerosols generation and characterization of effluents from oil shale processing were continued as part of an inhalation toxicology study. Additional data on plutonium excretion in urine are presented and point up problems in using the Langham equation to predict plutonium deposition in the body from long-term excretion data. Environmental surveillance at Los Alamos during 1983 showed the highest estimated radiation dose from Laboratory operations to be about 26% of the natural background radiation dose. Several studies on radionuclides and their transport in the Los Alamos environment are described. The chemical quality of surface and ground water near the geothermal hot dry rock facility is described. Short- and long-term consequences to man from releases of radionuclides into the environment can be simulated by the BIOTRAN computer model, which is discussed brirfly.

  11. New generation of NPP with boiling water reactor of improved safety

    SciTech Connect

    Adamov, E.O.; Kuklin, A.N.; Mityaev, Yu.I.; Mikhan, V.I.; Tokarev, Yu.I.; Cherkashov, Yu.M.; Sokolov, I.N.; Iljin, Yu.V.; Pakh, E.E.; Abramov, V.I.

    1993-12-31

    The nuclear power plants with boiling water reactors of improved safety are being developed. There is 26 years of operating experience with the plant VK-50 in Dimitrovgrad. The design and operation of the BWR reactors are described.

  12. Flibe use in fusion reactors -- An initial safety assessment

    SciTech Connect

    Cadwallader, L.C.; Longhurst, G.R.

    1999-03-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF{sub 2}) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  13. Flibe Use in Fusion Reactors - An Initial Safety Assessment

    SciTech Connect

    Cadwallader, Lee Charles; Longhurst, Glen Reed

    1999-04-01

    This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF2) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

  14. Safety characteristics of a suspended-pellet fission reactor system

    NASA Astrophysics Data System (ADS)

    Kingdon, David Ross

    A new fission reactor system with passive safety characteristics to eliminate the occurrence of loss-of-coolant accidents, reduce reactivity excursion effects, and which also provides for closure of the nuclear fuel cycle through on-site spent fuel management is examined. The concept uses multi-coated fuel pellets which are suspended by an upward moving coolant in vertical columns of the reactor core and electro-refining elemental separation to remove selected fission products prior to actinide recycling. The possibility of fuel melt following a loss-of-coolant is avoided as a decrease in coolant flow results in the removal of fuel from the core through the action of gravity alone. Average fluid velocities in the columns which are necessary to suspend the pellets are calculated and found to be consistent with the necessary heat extraction to yield ˜1--10 Wth per column. The total output power of such suspended pellet-type reactors is compared to the power necessary to provide the suspending fluid flow, yielding favourable ratios of ˜102--103. The reduction of reactivity excursion tendencies is envisaged through an ablative layer of material in the pellets which sublimates at temperatures above normal operating conditions. In the event of a power or temperature increase the particles fragment and thereby change their hydrodynamic drag characteristics, thus leading to fuel removal from the core by elutriation. Comparison of nuclear-to-thermal response times and elutriation rates for limiting power transients indicate that the present design assists in reactivity excursion mitigation. Closure of the nuclear fuel cycle is attained through a spent fuel management strategy which requires only on-site storage of a fraction of the fission products produced during reactor operation. Electro-refining separation of selected fission products combined with complete actinide recycling yields no isolation of plutonium or highly enriched uranium during the procedure. The out

  15. BESAFE II: Accident safety analysis code for MFE reactor designs

    NASA Astrophysics Data System (ADS)

    Sevigny, Lawrence Michael

    The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications

  16. A probabilistic safety analysis of incidents in nuclear research reactors.

    PubMed

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64. PMID:22021060

  17. Safety Characteristics of LBE Cooled Long-Life Small Reactor, 'LSPR'

    SciTech Connect

    Hiroshi Sekimoto; Shinichi Makino

    2002-07-01

    Lead bismuth eutectic (LBE) shows a good performance on neutron economy, and LBE cooled fast reactor can be designed as an excellent long-life small reactor. LBE is good not only for neutron economy but for chemical inertness and high boiling point, which may realize a much safer reactor than conventional sodium-cooled reactor. We have designed such a long-life small reactor and name it LSPR. This paper presents safety characteristics of LSPR. (authors)

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

    SciTech Connect

    Corradini, M. L.

    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

  19. Safety Issues at the DOE Test and Research Reactors. A Report to the U.S. Department of Energy.

    ERIC Educational Resources Information Center

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety issues at the Department of Energy (DOE) test and research reactors. Part A identifies six safety issues of the reactors. These issues include the safety design philosophy, the conduct of safety reviews, the performance of probabilistic risk assessments, the reliance on reactor operators, the fragmented…

  20. Safety of evolutionary and innovative nuclear reactors: IAEA activities and world efforts

    SciTech Connect

    Saito, T.; Gasparini, M.

    2004-07-01

    'Defence in Depth' approach constitutes the basis of the IAEA safety standards for nuclear power plants. Lessons learned from the current generation of reactors suggest that, for the next generation of reactor designs, the Defence in Depth philosophy should be retained, and that its implementation should be guided by the probabilistic insights. Recent developments in the area of general safety requirements based on Defence in Depth approach are examined and summarized. Global efforts to harmonize safety requirements for evolutionary nuclear power plants have involved many countries and organizations such as IAEA, US EPRI and European Utility EUR Organization. In recent years, developments of innovative nuclear power plants are also being discussed. The IAEA is currently developing a safety approach specifically for innovative nuclear reactors. This approach will eventually lead to a proposal of safety requirements for innovative reactors. Such activities related to safety requirements of evolutionary and innovative reactors are introduced. Various evolutionary and innovative reactor designs are reported in the world. The safety design features of evolutionary large LWRs, innovative LWRs, Modular High Temperature Gas Reactors and Small Liquid Metal Cooled LMRs are also introduced. Enhanced safety features proposed in such reactors are discussed and summarized according to the levels of Defence in Depth. For future nuclear plants, international cooperation and harmonization, especially in the area of safety, appear to be inevitable. Based on the past experience with many member states, the IAEA believes itself to be the uniquely positioned international organization to play this key role. (authors)

  1. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    SciTech Connect

    Carbajo, Juan; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Schmidt, Rodney Cannon; Thomas, Justin; Wei, Tom; Sofu, Tanju; Ludewig, Hans; Tobita, Yoshiharu; Ohshima, Hiroyuki; Serre, Frederic

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  2. Progress in reliability of fast reactor operation and new trends to increased inherent safety

    SciTech Connect

    Merk, Bruno; Stanculescu, Alexander; Chellapandi, Perumal; Hill, Robert

    2015-06-01

    The reasons for the renewed interest in fast reactors and an overview of the progress in sodium cooled fast reactor operation in the last ten years are given. The excellent operational performance of sodium cooled fast reactors in this period is highlighted as a sound basis for the development of new fast reactors. The operational performance of the BN-600 is compared and evaluated against the performance of German light water reactors to assess the reliability. The relevance of feedback effects for safe reactor design is described, and a new method for the enhancement of feedback effects in fast reactors is proposed. Experimental reactors demonstrating the inherent safety of advanced sodium cooled fast reactor designs are described and the potential safety improvements resulting from the use of fine distributed moderating material are discussed.

  3. Nuclear reactor safety research since three mile island.

    PubMed

    Mynatt, F R

    1982-04-01

    The Three Mile Island nuclear power plant accident has resulted in redirection of reactor safety research priorities. The small release to the environment of radioactive iodine-13 to 17 curies in a total radioactivity release of 2.4 million to 13 million curies-has led to a new emphasis on the physical chemistry of fission product behavior in accidents; the fact that the nuclear core was severely damaged but did not melt down has opened a new accident regime-that of the degraded core; the role of the operators in the progression and severity of the accident has shifted emphasis from equipment reliability to human reliability. As research progresses in these areas, the technical base for regulation and risk analysis will change substantially. PMID:17736229

  4. Very high temperature measurements: Application to nuclear reactor safety tests

    NASA Astrophysics Data System (ADS)

    Parga, Clemente Jose

    This PhD dissertation focuses on the improvement of very high temperature thermometry (1100ºC to 2480ºC), with special emphasis on the application to the field of nuclear reactor safety and severe accident research. Two main projects were undertaken to achieve this objective: -The development, testing and transposition of high-temperature fixed point (HTFP) metal-carbon eutectic cells, from metrology laboratory precision (+/-0.001ºC) to applied research with a reasonable degradation of uncertainties (+/-3-5ºC). -The corrosion study and metallurgical characterization of Type-C thermocouple (service temp. 2300ºC) prospective sheath material was undertaken to extend the survivability of TCs used for molten metallic/oxide corium thermometry (below 2000ºC).

  5. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    SciTech Connect

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

  6. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect

    Chad Pope

    2007-05-01

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day

  7. SACRD: a data base for fast reactor safety computer codes, operational procedures

    SciTech Connect

    Forsberg, V.M.; Arwood, J.W.; Greene, N.M.; Raiford, G.B.

    1980-09-01

    SACRD (Safety Analysis Computerized Reactor Data) is a data base of nondesign-related information used in computer codes for fast reactor safety analyses. This document reports the procedures used in SACRD to help assure a reasonable level of integrity of the material contained in the data base. It also serves to document much of the computer software used with the data base.

  8. Development of electro-optical instrumentation for reactor safety studies

    SciTech Connect

    Turko, B.T.; Kolbe, W.F.; Leskovar, B.; Sun, R.K.

    1980-11-01

    The development of new electro-optical instrumentation for reactor safety studies is described. The system measures the thickness of the water film and droplet size and velocity distributions which would be encountered in the annular two-phase flow in a reactor cooling system. The water film thickness is measured by a specially designed capacitance system with a short time constant. Water droplet size and velocity are measured by a subsystem consisting of a continuously pulsed laser light source, a vidicon camera, a video recorder, and an automatic image analyzer. An endoscope system attached to the video camera is used to image the droplets. Each frame is strobed with two accurately spaced uv light pulses, from two sequentially fired nitrogen lasers. The images are stored in the video disk recorder. The modified automatic image analyzer is programmed to digitize the droplet size and velocity distributions. Many special optical, mechanical and electronic system components were designed and fabricated. They are described in detail, together with calibration charts and experimental results.

  9. COUNTERCURRENT FLOW LIMITATION EXPERIMENTS AND MODELING FOR IMPROVED REACTOR SAFETY

    SciTech Connect

    Vierow, Karen

    2008-09-26

    This project is investigating countercurrent flow and “flooding” phenomena in light water reactor systems to improve reactor safety of current and future reactors. To better understand the occurrence of flooding in the surge line geometry of a PWR, two experimental programs were performed. In the first, a test facility with an acrylic test section provided visual data on flooding for air-water systems in large diameter tubes. This test section also allowed for development of techniques to form an annular liquid film along the inner surface of the “surge line” and other techniques which would be difficult to verify in an opaque test section. Based on experiences in the air-water testing and the improved understanding of flooding phenomena, two series of tests were conducted in a large-diameter, stainless steel test section. Air-water test results and steam-water test results were directly compared to note the effect of condensation. Results indicate that, as for smaller diameter tubes, the flooding phenomena is predominantly driven by the hydrodynamics. Tests with the test sections inclined were attempted but the annular film was easily disrupted. A theoretical model for steam venting from inclined tubes is proposed herein and validated against air-water data. Empirical correlations were proposed for air-water and steam-water data. Methods for developing analytical models of the air-water and steam-water systems are discussed, as is the applicability of the current data to the surge line conditions. This report documents the project results from July 1, 2005 through June 30, 2008.

  10. Lessons from Fukushima for Improving the Safety of Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Lyman, Edwin

    2012-02-01

    The March 2011 accident at the Fukushima Daiichi nuclear power plant has revealed serious vulnerabilities in the design, operation and regulation of nuclear power plants. While some aspects of the accident were plant- and site-specific, others have implications that are broadly applicable to the current generation of nuclear plants in operation around the world. Although many of the details of the accident progression and public health consequences are still unclear, there are a number of lessons that can already be drawn. The accident demonstrated the need at nuclear plants for robust, highly reliable backup power sources capable of functioning for many days in the event of a complete loss of primary off-site and on-site electrical power. It highlighted the importance of detailed planning for severe accident management that realistically evaluates the capabilities of personnel to carry out mitigation operations under extremely hazardous conditions. It showed how emergency plans rooted in the assumption that only one reactor at a multi-unit site would be likely to experience a crisis fail miserably in the event of an accident affecting multiple reactor units simultaneously. It revealed that alternate water injection following a severe accident could be needed for weeks or months, generating large volumes of contaminated water that must be contained. And it reinforced the grim lesson of Chernobyl: that a nuclear reactor accident could lead to widespread radioactive contamination with profound implications for public health, the economy and the environment. While many nations have re-examined their policies regarding nuclear power safety in the months following the accident, it remains to be seen to what extent the world will take the lessons of Fukushima seriously and make meaningful changes in time to avert another, and potentially even worse, nuclear catastrophe.

  11. Radioactive waste shipments to Hanford retrievable storage from Westinghouse Advanced Reactors and Nuclear Fuels Divisions, Cheswick, Pennsylvania

    SciTech Connect

    Duncan, D.; Pottmeyer, J.A.; Weyns, M.I.; Dicenso, K.D.; DeLorenzo, D.S.

    1994-04-01

    During the next two decades the transuranic (TRU) waste now stored in the burial trenches and storage facilities at the Hanford Sits in southeastern Washington State is to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico for final disposal. Approximately 5.7 percent of the TRU waste to be retrieved for shipment to WIPP was generated by the decontamination and decommissioning (D&D) of the Westinghouse Advanced Reactors Division (WARD) and the Westinghouse Nuclear Fuels Division (WNFD) in Cheswick, Pennsylvania and shipped to the Hanford Sits for storage. This report characterizes these radioactive solid wastes using process knowledge, existing records, and oral history interviews.

  12. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  13. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a...

  14. 75 FR 28200 - Safety Zone; Washington State Department of Transportation Ferries Division Marine Rescue...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-05-20

    ... Transportation Ferries Division Marine Rescue Response (M2R) Full-Scale Exercise for a Mass Rescue Incident (MRI... Madison. This training exercise will simulate a mass rescue incident (MRI) and will involve an abandon... Department of Transportation Ferries Division Marine Rescue Response (M2R) Full-Scale Exercise for a...

  15. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    SciTech Connect

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy`s (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors.

  16. Recent performance experience with US light water reactor self-actuating safety and relief valves

    SciTech Connect

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  17. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect

    Not Available

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24.

  18. Sodium fast reactor safety and licensing research plan. Volume I.

    SciTech Connect

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  19. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    SciTech Connect

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-07-01

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  20. Criticality safety analysis on fissile materials in Fukushima reactor cores

    SciTech Connect

    Liu, Xudong; Lemaitre-Xavier, E.; Ahn, Joonhong; Hirano, Fumio

    2013-07-01

    The present study focuses on the criticality analysis for geological disposal of damaged fuels from Fukushima reactor cores. Starting from the basic understanding of behaviors of plutonium and uranium, a scenario sequence for criticality event is considered. Due to the different mobility of plutonium and uranium in geological formations, the criticality safety is considered in two parts: (1) near-field plutonium system and (2) far-field low enriched uranium (LEU) system. For the near-field plutonium system, a mathematical analysis for pure-solute transport was given, assuming a particular buffer material and waste form configuration. With the transport and decay of plutonium accounted, the critical mass of plutonium was compared with the initial load of a single canister. Our calculation leads us to the conclusion that our system with the initial loading being the average mass of plutonium in an assembly just before the accident is very unlikely to become critical over time. For the far-field LEU system, due to the uncertainties in the geological and geochemical conditions, calculations were made in a parametric space that covers the variation of material compositions and different geometries. Results show that the LEU system could not remain sub-critical within the entire parameter space assumed, although in the iron-rich rock, the neutron multiplicity is significantly reduced.

  1. Passive and inherent safety technologies for light-water nuclear reactors

    SciTech Connect

    Forsberg, C.W.

    1990-07-01

    Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs.

  2. GIF sodium fast reactor project R and D on safety and operation

    SciTech Connect

    Vasile, A.; Sofu, T.; Jeong, H. Y.; Sakai, T.

    2012-07-01

    The 'Safety and Operation' project is started in 2009 within the framework of Generation-IV International Forum (GIF) Sodium Fast Reactor (SFR) research and development program. In the safety area, the project involves R and D activities on phenomenological model development and experimental programs, conceptual studies in support of the design of safety provisions, preliminary assessment of safety systems, framework and methods for analysis of safety architecture. In the operation area, the project involves R and D activities on fast reactors safety tests and analysis of reactor operations, feedback from decommissioning, in-service inspection technique development, under-sodium viewing and sodium chemistry. This paper presents a summary of such activities and the main achievements. (authors)

  3. Fast reactor safety: proceedings of the international topical meeting. Volume 1

    SciTech Connect

    Not Available

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 1 include: impact of safety and licensing considerations on fast reactor design; safety aspects of innovative designs; intra-subassembly behavior; operational safety; design accommodation of seismic and other external events; natural circulation; safety design concepts; safety implications derived from operational plant data; decay heat removal; and assessment of HCDA consequences.

  4. EH&S annual report: Summary of activities Environment, Health and Safety Division, 1992

    SciTech Connect

    Not Available

    1994-03-01

    This report presents an overview of the environment, safety, and health program in operation at the Lawrence Berkeley Laboratory. description of research in environmental science, remediation, waste management, safety, health services, radiation assessment, and emergency plans are provided.

  5. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  6. Safety Analyses at the Idaho National Engineering and Environmental Laboratory Test Reactor Area - Past to Present

    SciTech Connect

    Ambrosek, Richard Garry; Ingram, Frederick William

    1999-11-01

    Test reactors are unique in that the core configuration may change with each operating interval. The process of safety analyses for test reactors at the Idaho National Engineering and Environmental Test Reactor Area has evolved as the computing capabilities, software, and regulatory requirements have changed. The evaluations for experiments and the reactor have moved from measurements in a set configuration and then application to other configurations with a relatively large error to modeling in three-dimensions and explicit analyses for each experiment and operating interval. This evolution is briefly discussed for the Test Reactor Area.

  7. A new safety channel based on ¹⁷N detection in research reactors.

    PubMed

    Seyfi, Somayye; Gharib, Morteza

    2015-10-01

    Tehran research reactor (TRR) is a representative of pool type research reactors using light water, as coolant and moderator. This reactor is chosen as a prototype to demonstrate and prove the feasibility of (17)N detection as a new redundant channel for reactor power measurement. In TRR, similar to other pool type reactors, neutron detectors are immersed in the pool around the core as the main power measuring devices. In the present article, a different approach, using out of water neutron detector, is employed to measure reactor power. This new method is based on (17)O (n,p) (17)N reaction taking place inside the core and subsequent measurement of delayed neutrons emitted due to (17)N disintegration. Count and measurement of neutrons around outlet water pipe provides a reliable redundant safety channel to measure reactor power. Results compared with other established channels indicate a good agreement and shows a linear interdependency with true thermal power. Safety of reactor operation is improved with installation & use of this new power measuring channel. The new approach may equally serve well as a redundant channel in all other types of reactors having coolant comprised of oxygen in its molecular constituents. Contrary to existing channels, this one is totally out of water and thus is an advantage over current instrumentations. It is proposed to employ the same idea on other reactors (nuclear power plants too) to improve safety criteria. PMID:26123105

  8. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, C.W.

    1985-02-19

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (nonborated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two water volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  9. Boiling water neutronic reactor incorporating a process inherent safety design

    DOEpatents

    Forsberg, Charles W.

    1987-01-01

    A boiling-water reactor core is positioned within a prestressed concrete reactor vessel of a size which will hold a supply of coolant water sufficient to submerge and cool the reactor core by boiling for a period of at least one week after shutdown. Separate volumes of hot, clean (non-borated) water for cooling during normal operation and cool highly borated water for emergency cooling and reactor shutdown are separated by an insulated wall during normal reactor operation with contact between the two water volumes being maintained at interfaces near the top and bottom ends of the reactor vessel. Means are provided for balancing the pressure of the two volumes at the lower interface zone during normal operation to prevent entry of the cool borated water into the reactor core region, for detecting the onset of excessive power to coolant flow conditions in the reactor core and for detecting low water levels of reactor coolant. Cool borated water is permitted to flow into the reactor core when low reactor coolant levels or excessive power to coolant flow conditions are encountered.

  10. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

    SciTech Connect

    Ho, S.K.; Fowler, T.K.; Holdren, J.P.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  11. Student research activities in the Technology Assessments Section of the Health and Safety Research Division, Summer 1980

    SciTech Connect

    Chester, R.O.; Roberts, D.A.

    1981-08-01

    Reports summarizing activities of students assigned to the Technology Assessments Section of the Health and Safety Research Division for the summer 1980 are presented. Unless indicated otherwise, each report was written by the student whose work is being described. For each student, the student's supervisor, the name of the program under which the student was brought to ORNL, the academic level of the student, and the name of the ORNL project to which the student was assigned are tabulated. The reports are presented in alphabetical order of the students' last names.

  12. The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

    SciTech Connect

    Hurt, S.S.

    1986-11-01

    The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary.

  13. A new concept of nuclear fission reactors safety

    SciTech Connect

    Petrov, Y.V.

    1993-12-31

    To develop safe nuclear energy production acceptable to the society it is proposed to use in the future strongly subcritical reactors (k=0.96-0.97) driven by proton or deuteron accelerators. The accelerator with the current of 40mA and particle energy {approximately}0.8 GeV/nucleon will provide 2 GW (th.) reactor power in fast reactor with metallic U-Pu fuel. The design, control and parameters of such a system are discussed.

  14. Industrial Safety and Applied Health Physics Division, annual report for 1982

    SciTech Connect

    Not Available

    1983-12-01

    Activities during the past year are summarized for the Health Physics Department, the Environmental Management Department, and the Safety Department. The Health Physics Department conducts radiation and safety surveys, provides personnel monitoring services for both external and internal radiation, and procures, services, and calibrates appropriate portable and stationary health physics instruments. The Environmental Management Department insures that the activities of the various organizations within ORNL are carried out in a responsible and safe manner. This responsibility involves the measurement, field monitoring, and evaluation of the amounts of radionuclides and hazardous materials released to the environment and the control of hazardous materials used within ORNL. The department also collaborates in the design of ORNL Facilities to help reduce the level of materials released to the environment. The Safety Department is responsible for maintaining a high level of staff safety. This includes aspects of both operational and industrial safety and also coordinates the activities of the Director's Safety Review Committee. (ACR)

  15. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  16. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-12

    ... COMMISSION Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES:...

  17. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  18. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  19. THE RADIATION SAFETY INFORMATION COMPUTATIONAL CENTER: A RESOURCE FOR REACTOR DOSIMETRY SOFTWARE AND NUCLEAR DATA

    SciTech Connect

    Kirk, Bernadette Lugue

    2009-01-01

    The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

  20. The Radiation Safety Information Computational Center:. a Resource for Reactor Dosimetry Software and Nuclear Data

    NASA Astrophysics Data System (ADS)

    Kirk, B. L.

    2009-08-01

    The Radiation Safety Information Computational Center (RSICC) was established in 1963 to collect and disseminate computational nuclear technology in the form of radiation transport, shielding and safety software and corresponding nuclear cross sections. Approximately 1700 nuclear software and data packages are in the RSICC collection, and the majority are applicable to reactor dosimetry.

  1. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.58 Safety/security...

  2. Safety aspect concerning radiolytic gas generation in reactors.

    PubMed

    Ramshesh, V

    2001-01-01

    In water cooled and water moderated reactors (H2O in boiling water reactors/pressurised water reactors, D2O in pressurised heavy water reactors) during normal operation, radiolysis is a source of production of hydrogen/deuterium and oxygen. During the progress of a nuclear accident, while there are other important sources of hydrogen/deuterium, the oxygen availability can occur only through radiolysis or direct contact with air. In air saturated with water vapour at room temperature and pressure when H2/D2 concentration exceeds 4 vol % (a conservative estimate), a combustible mixture with oxygen can be formed. It is proposed to examine the basic principles of water radiolysis as far as they pertain to generation of H2/D2 and O2 and try to apply these concepts to reactors both under operating conditions and in accident situations. It is concluded that the possibility of an accident taking place through radiolysis is highly unlikely. PMID:11382138

  3. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    SciTech Connect

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  4. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. PMID:21399407

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

    SciTech Connect

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

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

  6. Developing a Comprehensive Software Suite for Advanced Reactor Performance and Safety Analysis

    SciTech Connect

    Pointer, William David; Bradley, Keith S; Fischer, Paul F; Smith, Micheal A; Tautges, Timothy J; Ferencz, Robert M; Martineau, Richard C; Jain, Rajeev; Obabko, Aleksandr; Billings, Jay Jay

    2013-01-01

    This paper provides an introduction to the reactor analysis capabilities of the nuclear power reactor simulation tools that are being developed as part of the U.S. Department of Energy s Nuclear Energy Advanced Modeling and Simulation (NEAMS) Toolkit. The NEAMS Toolkit is an integrated suite of multi-physics simulation tools that leverage high-performance computing to reduce uncertainty in the prediction of performance and safety of advanced reactor and fuel designs. The Toolkit effort is comprised of two major components, the Fuels Product Line (FPL), which provides tools for fuel performance analysis, and the Reactor Product Line (RPL), which provides tools for reactor performance and safety analysis. This paper provides an overview of the NEAMS RPL development effort.

  7. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  8. Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

    SciTech Connect

    Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

    1984-09-01

    The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits.

  9. Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

    SciTech Connect

    Cook, David Howard

    2009-01-01

    A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities.

  10. Safety Issues at the Defense Production Reactors. A Report to the U.S. Department of Energy.

    ERIC Educational Resources Information Center

    National Academy of Sciences - National Research Council, Washington, DC. Commission on Physical Sciences, Mathematics, and Resources.

    This report provides an assessment of safety management, safety review, and safety methodology employed by the Department of Energy (DOE) and private contractors. Chapter 1, "The DOE Safety Framework," examines safety objectives for production reactors and processes to implement the objectives. Chapter 2, "Technical Issues," focuses on a variety…

  11. Simulator platform for fast reactor operation and safety technology demonstration

    SciTech Connect

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

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

    SciTech Connect

    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.

  13. Technology Development, Evaluation, and Application (TDEA) FY 1995 progress report - Environmental, Safety, and Health (ESH) division

    SciTech Connect

    Andrews, L.L.

    1996-09-01

    This report covers six months of effort, including startup time. Five projects were supported by the division: Pilot Program for the Risk-Based Surveillance of Lung Cancer in Los Alamos National Laboratory Workers, Optimization of Placement of Workplace Continuous Air Monitoring Instrumentation, A Polymeric Barrier Monitor to Protect Workers, Evaluation of a Real-Time Beryllium Detection Instrument and the Implications of Its Use, and High-Energy Dosimetry. A project summary for each is provided. An appendix to the report includes the 1995 Request for Proposals, Committee Members, Priority Technical Areas of Interest for FY95, Relative Prioritization and Weighting Factors, Format for Proposals, and Charter.

  14. Recent developments in Topaz II reactor safety assessments

    SciTech Connect

    Marshall, A.C.

    1993-07-01

    In December 1991, the Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of a US launch of a Russian Topaz II space nuclear power system. The primary mission goal would be to demonstrate and evaluate Nuclear Electric Propulsion technology to establish a capability for future civilian and military missions. A preliminary nuclear safety assessment, involving selected safety analyses, was initiated to determine whether or not a space mission could be conducted safely and within budget constraints. This paper describes the preliminary safety assessment results and the nuclear safety program now being established for the Nuclear Electric Propulsion Space Test Program (NEPSTP).

  15. Health and Safety Research Division: Progress report, October 1, 1985-March 31, 1987

    SciTech Connect

    Walsh, P.J.

    1987-09-01

    This report summarizes the progress in our programs for the period October 1, 1985, through March 31, 1987. The division's presentations and publications represented important contributions on the forefronts of many fields. Eleven invention disclosures were filed, two patent applications submitted, and one patent issued. The company's transfers new technologies to the private sector more efficiently than in the past. The division's responsibilities to DOE under the Uranium Mill Tailings Remedial Action (UMTRA) program includes inclusion recommendations for 3100 properties. The nuclear medicine program developed new radiopharmaceuticals and radionuclide generators through clinical trials with some of our medical cooperatives. Two major collaborative indoor air quality studies and a large epidemiological study of drinking water quality and human health were completed. ORNL's first scanning tunneling microscope (STM) has achieved single atom resolution and has produced some of the world's best images of single atoms on the surface of a silicon crystal. The Biological and Radiation Physics Section, designed and constructed a soft x-ray spectrometer which has exhibited a measuring efficiency that is 10,000 times higher than other equipment. 1164 refs.

  16. Advanced reactor safety program. Stakeholder interaction and feedback

    SciTech Connect

    Szilard, Ronaldo H.; Smith, Curtis L.

    2014-08-01

    In the Spring of 2013, we began discussions with our industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require substantial participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

  17. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    SciTech Connect

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  18. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    SciTech Connect

    1983-10-01

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  19. Recent developments in Topaz-II reactor safety assessments

    SciTech Connect

    Marshall, A.C. )

    1993-01-01

    In December 1991, the Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of a US launch of a Russian Topaz-II space nuclear power system. The primary mission goal would be to demonstrate and evaluate nuclear electric propulsion technology to establish a capability for future civilian and military missions. A preliminary nuclear safety analysis was initiated to determine whether or not a space mission could be conducted safely and within budget constraints. This paper describes preliminary safety analysis results and the nuclear safety program now being established for the NEP space test (NEPST).

  20. Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

    SciTech Connect

    Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao

    2008-09-01

    This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un’SCRAM’ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role.

  1. Preliminary safety calculations to improve the design of Molten Salt Fast Reactor

    SciTech Connect

    Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A.

    2012-07-01

    Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

  2. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    SciTech Connect

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  3. Neutronic safety parameters and transient analyses for potential LEU conversion of the Budapest Research Reactor.

    SciTech Connect

    Pond, R. B.; Hanan, N. A.; Matos, J. E.; Maraczy, C.

    1999-09-27

    An initial safety study for potential LEU conversion of the Budapest Research Reactor was completed. The study compares safety parameters and example transients for reactor cores with HEU and LEU fuels. Reactivity coefficients, kinetic parameters and control rod worths were calculated for cores with HEU(36%) UAl alloy fuel and UO{sub 2}-Al dispersion fuel, and with LEU (19.75%)UO{sub 2}-Al dispersion fuel that has a uranium density of about 2.5 g/cm{sup 3}. A preliminary fuel conversion plan was developed for transition cores that would convert the BRR from HEU to LEU fuel after the process is begun.

  4. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    SciTech Connect

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant.

  5. Evaluation of Launch Accident Safety Options for Low-Power Surface Reactors

    SciTech Connect

    Fung Poon, Cindy; Poston, David I.

    2006-01-20

    Safety options for surface reactors of less than 800 kW (thermal power) are analyzed. The concepts under consideration are heat pipe cooled reactors fueled with either uranium nitride or uranium dioxide. This study investigates the impact of launch accident criteria on the system mass, while ensuring the mechanical integrity and reliability of the system through launch accident scenarios. The four criticality scenarios analyzed for shutdown determination are dry sand surround with reflectors stripped, water submersion on concrete, water submersion with all control drums in, and the nominal shutdown reactor condition. Additionally the following two operational criteria are analyzed: reactor is warm and swelled, and reactor is warm and swelled with one drum in (where swelled includes both thermal mechanical expansion and irradiation induced swelling of the fuel)

  6. Transactions of the Twenty-First Water Reactor Safety Information Meeting

    SciTech Connect

    Monteleone, S.

    1993-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

  7. Transactions of the twenty-fifth water reactor safety information meeting

    SciTech Connect

    Monteleone, S.

    1997-09-01

    This report contains summaries of papers on reactor safety research to be presented at the 25th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 20--22, 1997. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion of information exchanged during the course of the meeting, and are given in order of their presentation in each session.

  8. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect

    Samuel Bays; Ayodeji Alajo

    2010-05-01

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  9. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    SciTech Connect

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  10. The uses and benefits of probabilistic risk assessment in nuclear reactor safety

    SciTech Connect

    Bari, R.A.; Speis, T.P.; Nuclear Regulatory Commission, Washington, DC . Office of Nuclear Regulatory Research)

    1989-01-01

    Probabilistic risk assessment (PRA) has proven to be an important tool in the safety assessment of nuclear reactors throughout the world. Decision making with regard to many safety issues has been facilitated by both general insights from and direct application of this technology. Key uses of PRA are discussed and some examples of successful applications are cited. The benefits and limitations of PRA are also discussed as well as the broader outlook for applications of PRA. 9 refs.

  11. Reactor safety research programs. Quarterly report, July-September 1983

    SciTech Connect

    Edler, S.K.

    1984-04-01

    Evaluations of nondestructive examination (NDE) techniques and instrumentation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, and examining NDE reliability and probabilistic fracture mechanics. Accelerated pellet-cladding interaction modeling is being conducted to predict the probability of fuel rod failure under normal operating conditions. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Experimental data and validated models are being used to determine a method for evaluating the acceptance of welded or weld-repaired stainless steel piping. Thermal-hydraulic models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada; and an instrumented fuel assembly irradiation program is being performed at Halden, Norway. Fuel assemblies and analytical support are being provided for experimental programs at other facilities, including the Super Sara Test Program, Ispra, Italy, and experimental programs at the Power Burst Facility.

  12. Safety goals and functional performance criteria. [Advanced Reactor Design

    SciTech Connect

    Youngblood, R.W.; Pratt, W.T. ); Hardin, W.B. . Div. of Regulatory Applications)

    1990-01-01

    This report discusses a possible approach to the development of functional performance criteria to be applied to evolutionary LWR designs. Key safety functions are first identified; then, criteria are drawn up for each individual function, based on the premise that no single function's projected unreliability should be allowed to exhaust the safety goal frequencies. In the area of core damage prevention, functional criteria are cast in terms of necessary levels of redundancy and diversity of critical equipment. In the area of core damage mitigation (containment), functional performance criteria are cast with the aim of mitigating post-core-melt phenomena with sufficient assurance to eliminate major uncertainties in containment performance. 9 refs.

  13. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-10

    ... solicitation for public comment published in the Federal Register on October 12, 2012 (77 FR 62270), on the... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks...

  14. Site Environmental Report for 2006. Volume I, Environment, Health, and Safety Division

    SciTech Connect

    2007-09-30

    Each year, Ernest Orlando Lawrence Berkeley National Laboratory prepares an integrated report on its environmental programs to satisfy the requirements of United States Department of Energy Order 231.1A, Environment, Safety, and Health Reporting.1 The Site Environmental Report for 2006 summarizes Berkeley Lab’s environmental management performance, presents environmental monitoring results, and describes significant programs for calendar year 2006. (Throughout this report, Ernest Orlando Lawrence Berkeley National Laboratory is referred to as “Berkeley Lab,” “the Laboratory,” “Lawrence Berkeley National Laboratory,” and “LBNL.”) The report is separated into two volumes. Volume I is organized into an executive summary followed by six chapters that contain an overview of the Laboratory, a discussion of the Laboratory’s environmental management system, the status of environmental programs, and summarized results from surveillance and monitoring activities. Volume II contains individual data results from surveillance and monitoring activities.

  15. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  16. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    SciTech Connect

    Klein, Andrew; Matthews, Topher; Lenhof, Renae; Deason, Wesley; Harter, Jackson

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  17. Magnetohydrodynamic Vapor Explosions: A Study with Potential Interest to the Safety of Fusion Reactor Project

    NASA Astrophysics Data System (ADS)

    Arias, F. J.

    2010-04-01

    In this paper, the possibility of vapor explosions in superheat liquids in presence of a magnetic field that undergo sudden variation of magnetic field is discussed. This possible phenomenon may play a very important role in the blanket design of future fusion reactors, where transients in magnetic field on liquid metals, could will be a potential hazard for safety.

  18. Two-field and drift-flux models with applications to nuclear reactor safety

    SciTech Connect

    Travis, J.R.

    1984-05-01

    The ideas of the two-field (6 equation model) and drift-flux (4 equation model) description of two-phase flows are presented. Several example calculations relating to reactor safety are discussed and comparisons of the numerical results and experimental data are shown to be in good agreement.

  19. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    SciTech Connect

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

    2004-12-31

    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  20. Inherent safety of minimum-burnup breed and burn reactors

    SciTech Connect

    Qvist, S.; Reenspan, E.

    2012-07-01

    Reactors that aim to sustain the breed and burn (B and B) mode of operation at minimum discharge burnup require excellent neutron economy, Minimum-burnup B and B cores are generally large and feature low neutron leakage probability and a hard neutron spectrum. While highly promising fuel cycles can be achieved with such designs, the very same features are pushing the limits of the core's ability to passively respond safely to unprotected accidents. Low leakage minimum-burnup sodium-cooled B and B cores have a large positive coolant void-worth and coolant temperature reactivity coefficient. In this study, the applicability of major approaches for fast reactor void-worth reduction is evaluated specifically for B and B cores. The design, shuffling scheme and performance of a new metallic-fueled, sodium-cooled minimum burnup B and B core, used as basis for the void-worth reduction analysis, is presented. The analysis shows that reactivity control systems based on passive {sup 6}Li injection during temperature excursions are the only option able to provide negative void-worth without significantly increasing the minimum burnup required for sustaining the B and B mode of operation. A new type of lithium expansion module (LEM) system was developed specifically for B and B cores and its effect on core performance is presented. (authors)

  1. Monte Carlo verification of point kinetics for safety analysis of nuclear reactors

    SciTech Connect

    Valentine, T.E.; Mihalczo, J.T.

    1995-06-01

    Monte Carlo neutron transport methods can be used to verify the applicability of point kinetics for safety analysis of nuclear reactors. KENO-NR was used to obtain the transfer function of the Advanced Neutron Source reactor and the time delay between the core power production and the external detectors, a parameter of interest to the safety systems design. The good agreement between the Monte Carlo generated transfer function and the point kinetics transfer function validates that the uncommon ANS geometry does not preclude the use of point kinetics in the frequency range that was investigated. Various features of the power spectral densities also demonstrated the applicability of point kinetics. The time delay was obtained from the cross-power spectral density (CPSD) and is {approximately}15 ms. These analyses show that frequency analysis can be used experimentally to investigate the validity of the use of point kinetics models in critical experiments or zero power testing of reactors.

  2. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    SciTech Connect

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  3. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    SciTech Connect

    Charak, I.; Kier, P.H.

    1995-04-01

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule ({section}50.62); (2) station blackout ({section}50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term ({section}50.34(f) and {section}100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements ({section}50.55a, etc); (6) ECCS acceptance criteria ({section}50.46)(b); (7) combustible gas control ({section}50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle ({section}51.51); and (11) (standards {section}50.55a).

  4. Site Environmental Report for 2004. Volume 1, Environment, Health, and Safety Division

    SciTech Connect

    2005-09-30

    Each year, Ernest Orlando Lawrence Berkeley National Laboratory prepares an integrated report on its environmental programs to satisfy the requirements of United States Department of Energy Order 231.1A, Environment, Safety, and Health Reporting.1 The Site Environmental Report for 2004 summarizes Berkeley Lab’s environmental management performance, presents environmental monitoring results, and describes significant programs for calendar year 2004. (Throughout this report, Ernest Orlando Lawrence Berkeley National Laboratory is referred to as “Berkeley Lab,” “the Laboratory,” “Lawrence Berkeley National Laboratory,” and “LBNL.”) The report is separated into two volumes. Volume I contains an overview of the Laboratory, the status of environmental programs, and summarized results from surveillance and monitoring activities. Volume II contains individual data results from these activities. This year, the Site Environmental Report was distributed by releasing it on the Web from the Berkeley Lab Environmental Services Group (ESG) home page, which is located at http://www.lbl.gov/ehs/esg/. Many of the documents cited in this report also are accessible from the ESG Web page. CD and printed copies of this Site Environmental Report are available upon request.

  5. Determination of core design thermal safety limits for a two-loop pressurized water reactor

    SciTech Connect

    Kostadinov, V.

    1996-04-01

    Results are given of independent research of core thermal design limits for the Nuklearna Elektrarna Krsko (NEK) nuclear power plant; procedures for two-loop pressurized water reactor plant core design safety limit calculation are used. Emphasis is placed on researching the vessel exit boiling and the hot-channel exit quality limits and their impact on the maximum available design safety operating range and thermal operating margin of the NEK reactor core. For this purpose, the LIMITS computer code is developed. Based on the modified, well-tried COBRA-IV-I computer code, the departure of nuclear boiling ratio core safety limits are calculated. The original results complement well those of the NEK Final Safety Analysis Report. The procedures and the methods for determining the reactor core design thermal limits are successfully proven despite the unavailability of proprietary data, different models, and computer codes. In addition to the acquired capability of in-house independent checking of the vendor`s results, the bases are set for further independent analyses of the limiting safety system settings for the NEK core.

  6. RISMC advanced safety analysis working plan: FY2015 - FY2019. Light Water Reactor Sustainability Program

    SciTech Connect

    Szilard, Ronaldo H; Smith, Curtis L

    2014-09-01

    In this report, the Advanced Safety Analysis Program (ASAP) objectives and value proposition is described. ASAP focuses on modernization of nuclear power safety analysis (tools, methods and data); implementing state-of-the-art modeling techniques (which include, for example, enabling incorporation of more detailed physics as they become available); taking advantage of modern computing hardware; and combining probabilistic and mechanistic analyses to enable a risk informed safety analysis process. The modernized tools will maintain the current high level of safety in our nuclear power plant fleet, while providing an improved understanding of safety margins and the critical parameters that affect them. Thus, the set of tools will provide information to inform decisions on plant modifications, refurbishments, and surveillance programs, while improving economics. The set of tools will also benefit the design of new reactors, enhancing safety per unit cost of a nuclear plant. As part of the discussion, we have identified three sets of stakeholders, the nuclear industry, the Department of Energy (DOE), and associated oversight organizations. These three groups would benefit from ASAP in different ways. For example, within the DOE complex, the possible applications that are seen include the safety of experimental reactors, facility life extension, safety-by-design in future generation advanced reactors, and managing security for the storage of nuclear material. This report provides information in five areas: (1) A value proposition (“why is this important?”) that will make the case for stakeholder’s use of the ASAP research and development (R&D) products; (2) An identification of likely end users and pathway to adoption of enhanced tools by the end-users; (3) A proposed set of practical and achievable “use case” demonstrations; (4) A proposed plan to address ASAP verification and validation (V&V) needs; and (5) A proposed schedule for the multi-year ASAP.

  7. Tritium management in a fusion reactor--safety, handling and economical issues--

    SciTech Connect

    Tanabe, Tetsuo

    2009-02-19

    In order to establish a D-T fusion reactor as an energy source, it is not enough to have a DT burning plasma, and economical conversion of fusion energy to electricity and/or heat, a large enough margin of tritium breeding and tritium safety must be simultaneously achieved. In particular, handling of huge amount of tritium needs significant efforts to ensure that the radiation dose of radiological workers and of the public is below the limits specified by the International Commission on Radiological Protection (ICRP). In this paper, after the introduction of tritium as a fuel of DT reactors and as a radioisotope of hydrogen, tritium safety issues in fuel cycle and blanket systems are summarized. In particular, in-vessel tritium inventory, the most important and uncertain tritium safety issue, is discussed in detail.

  8. Fast reactor safety: proceedings of the international topical meeting. Volume 2. [R

    SciTech Connect

    Not Available

    1985-07-01

    The emphasis of this meeting was on the safety-related aspects of fast reactor design, analysis, licensing, construction, and operation. Relative to past meetings, there was less emphasis on the scientific and technological basis for accident assessment. Because of its broad scope, the meeting attracted 217 attendees from a wide cross section of the design, safety analysis, and safety technology communities. Eight countries and two international organizations were represented. A total of 126 papers were presented, with contributions from the United States, France, Japan, the United Kingdom, Germany, and Italy. Sessions covered in Volume 2 include: safety design concepts; operational transient experiments; analysis of seismic and external events; HCDA-related codes, analysis, and experiments; sodium fires; instrumentation and control/PPS design; whole-core accident analysis codes; and impact of safety design considerations on future LMFBR developments.

  9. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed

    2009-11-01

    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.

  10. Reviewing real-time performance of nuclear reactor safety systems

    SciTech Connect

    Preckshot, G.G.

    1993-08-01

    The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems.

  11. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    SciTech Connect

    Zhao, H.; Zhang, H.; Zou, L.; Sun, X.

    2012-07-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The RVACS can be characterized as a surface-based decay heat removal system. It is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to the core volume) and decay heat removal capability (proportional to the vessel surface area). Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  12. Generic safety insights for inspection of boiling water reactors

    SciTech Connect

    Higgins, J.C.; Taylor, J.H.; Fresco, A.N.; Hillman, B.M.

    1987-01-01

    As the number of operating nuclear power plants (NPP) increases, safety inspection has increased in importance. However, precisely what is important, and what is not important. What should one focus inspection efforts on. Over the last two years Probabilistic Risk Assessment (PR) techniques have been developed to aid in the inspection process. Broad interest in generic PRA-based methods has arisen in the past year, since only about 25% of the US nuclear power plants have completed PRAs, and also, inspectors want PRA-based tools for these plants. This paper describes the BNL program to develop generic BWR PRA-based inspection insights or inspection guidance designed to be applied to plants without PRAs.

  13. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    SciTech Connect

    Tomberlin, Terry Alan

    2002-06-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to "major modifications" and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

  14. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    SciTech Connect

    Tomberlin, T.A.

    2002-06-19

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed.

  15. Energy Technology Division research summary -- 1994

    SciTech Connect

    Not Available

    1994-09-01

    Research funded primarily by the NRC is directed toward assessing the roles of cyclic fatigue, intergranular stress corrosion cracking, and irradiation-assisted stress corrosion cracking on failures in light water reactor (LWR) piping systems, pressure vessels, and various core components. In support of the fast reactor program, the Division has responsibility for fuel-performance modeling and irradiation testing. The Division has major responsibilities in several design areas of the proposed International Thermonuclear Experimental Reactor (ITER). The Division supports the DOE in ensuring safe shipment of nuclear materials by providing extensive review of the Safety Analysis Reports for Packaging (SARPs). Finally, in the nuclear area they are investigating the safe disposal of spent fuel and waste. In work funded by DOE`s Energy Efficiency and Renewable Energy, the high-temperature superconductivity program continues to be a major focal point for industrial interactions. Coatings and lubricants developed in the division`s Tribology Section are intended for use in transportation systems of the future. Continuous fiber ceramic composites are being developed for high-performance heat engines. Nondestructive testing techniques are being developed to evaluate fiber distribution and to detect flaws. A wide variety of coatings for corrosion protection of metal alloys are being studied. These can increase lifetimes significant in a wide variety of coal combustion and gasification environments.

  16. Calculational framework for safety analyses of non-reactor nuclear facilities

    SciTech Connect

    Coleman, J.R.

    1994-06-01

    A calculational framework for the consequences analysis of non-reactor nuclear facilities is presented. The analysis framework starts with accident scenarios which are developed through a traditional hazard analysis and continues with a probabilistic framework for the consequences analysis. The framework encourages the use of response continua derived from engineering judgment and traditional deterministic engineering analyses. The general approach consists of dividing the overall problem into a series of interrelated analysis cells and then devising Markov chain like probability transition matrices for each of the cells. An advantage of this division of the problem is that intermediate output (as probability state vectors) are generated at each calculational interface. The series of analyses when combined yield risk analysis output. The analysis approach is illustrated through application to two non-reactor nuclear analyses: the Ulysses Space Mission, and a hydrogen burn in the Hanford waste storage tanks.

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

    NASA Astrophysics Data System (ADS)

    Day, Simon E.

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

  18. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    NASA Astrophysics Data System (ADS)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

  19. Dynamics and inherent safety features of small modular high temperature gas-cooled reactors

    SciTech Connect

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1986-01-01

    Investigations were made at Oak Ridge National Laboratory to characterize the dynamics and inherent safety features of various modular high temperature gas-cooled reactor (HTGR) designs. This work was sponsored by the US Nuclear Regulatory Commission's HTGR Safety Research program. The US Department of Energy (DOE) and the Gas Cooled Reactor Associates (GCRA) have sponsored studies of several modular HTGR concepts, each having it own unique advantageous economic and inherent safety features. The DOE design team has recently choses a 350-MW(t) annular core with prismatic, graphite matrix fuel for its reference plant. The various safety features of this plant and of the pebble-bed core designs similar to those currently being developed and operated in the Federal Republic of Germany (FRG) are described. A varity of postulated accident sequences involving combinations of loss of forced circulation of the helium primary coolant, loss of primary coolant pressurization, and loss of normal and backup heat sinks were studied and are discussed. Results demonstrate that each concept can withstand an uncontrolled heatup accident without reaching excessive peak fuel temperatures. Comparisons of calculated and measured response for a loss of forced circulation test on the FRG reactor, AVR, are also presented. 10 refs.

  20. Reactor safety research programs. Quarterly report, April-June 1983. Vol. 2

    SciTech Connect

    Edler, S.K.

    1983-12-01

    This document summarizes work performed by Pacific Northwest Laboratory from April 1 through June 30, 1983, for the Division of Accident Evaluation and the Division of Engineering Technology, US Nuclear Regulatory Commission. Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Experimental data and validated models are being used to determine a method for evaluating the acceptance of welded or weld-repaired stainless steel piping. Core thermal models are being developed to provide better digital codes to compute the behavior or full-scale reactor systems under postulated accident conditions. High-temperature materials property tests are being conducted to provide data on severe core damage fuel behavior. Severe fuel damage accident tests are being conducted at the NRU reactor, Chalk River, Canada; and an instrumented fuel assembly irradiation program is being performed at Halden, Norway. Fuel assemblies and analytical support are being provided for experimental programs at other facilities, including fuel rod deformation and severe fuel damage tests for the Super Sara Test Program, Ispra, Italy; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory, Idaho Falls, Idaho.

  1. Critical review of the reactor-safety study radiological health effects model. Final report

    SciTech Connect

    Cooper, D.W.; Evans, J.S.; Jacob, N.; Kase, K.R.; Maletskos, C.J.; Robertson, J.B.; Smith, D.G.

    1983-03-01

    This review of the radiological health effects models originally presented in the Reactor Safety Study (RSS) and currently used by the US Nuclear Regulatory Commission (NRC) was undertaken to assist the NRC in determining whether or not to revise the models and to aid in the revision, if undertaken. The models as presented in the RSS and as implemented in the CRAC (Calculations of Reactor Accident Consequences) Code are described and critiqued. The major elements analyzed are those concerning dosimetry, early effects, and late effects. The published comments on the models are summarized, as are the important findings since the publication of the RSS.

  2. Final report-passive safety optimization in liquid sodium-cooled reactors.

    SciTech Connect

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  3. An approach to model reactor core nodalization for deterministic safety analysis

    NASA Astrophysics Data System (ADS)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  4. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    SciTech Connect

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  5. (Instrumentation and controls technology and reactor operational safety)

    SciTech Connect

    White, J.D.

    1990-12-17

    While on vacation, the traveler participated as a co-chairman of a panel of instrumentation and controls specialists visiting nuclear establishments in Europe. The purpose of the visit was to assess the status of instrumentation and controls technology for nuclear power in Europe. A list of the sites visited and the personnel contacted is included in this trip report. The visit was sponsored by Loyola College working under contract to the National Science Foundation. All costs were paid by Loyola College, for whom the traveler was a consultant. This was an outside activity approved by DOE. The traveler was surprised by the high level of automaton present in the German Konvoi nuclear power plants built by Siemens AG KWU. The claim was that this was done to improve the safety of the plant by keeping the operator out of the loop'' for the first 30 minutes of some transients or accidents. The traveler was also surprised by the high level of man-machine interface R D in the USSR.

  6. Plant Modernization with Digital Reactor Protection System Safety System Upgrades at US Nuclear Power Stations

    SciTech Connect

    Heckle, Wm. Lloyd; Bolian, Tricia W.

    2006-07-01

    As the current fleet of nuclear power plants in the US reaches 25+ years of operation, obsolescence is driving many utilities to implement upgrades to both their safety and non-safety-related Instrumentation and Control (I and C) Systems. Digital technology is the predominant replacement technology for these upgrades. Within the last 15 years, digital control systems have been deployed in non-safety- related control applications at many utilities. In addition, a few utilities have replaced small safety-related systems utilizing digital technology. These systems have shown digital technology to be robust, reliable and simpler to maintain. Based upon this success, acceptance of digital technology has gained momentum with both utilities and regulatory agencies. Today, in an effort to extend the operating lives of their nuclear stations and resolve obsolescence of critical components, utilities are now pursuing digital technology for replacement of their primary safety systems. AREVA is leading this effort in the United States with the first significant digital upgrade of a major safety system. AREVA has previously completed upgrades to safety-related control systems emergency diesel engine controls and governor control systems for a hydro station which serves as the emergency power source for a nuclear station. Currently, AREVA is implementing the replacement of both the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS) on all three units at a US PWR site. (authors)

  7. Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

    SciTech Connect

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  8. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    SciTech Connect

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  9. Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project

    SciTech Connect

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  10. Materials and mechanical design analysis of boron carbide reactor safety rods. Final report

    SciTech Connect

    Marra, J.C.

    1992-04-01

    The purpose of this task was to analyze the materials and mechanical design bases for the new boron carbide safety rod. These analyses included examination of the irradiation response of the materials, chemical compatibility of component materials, moisture considerations for the boron carbide pellets and susceptibility of the rod to corrosion under reactor environmental conditions. A number of issues concerning the mechanical behavior were also addressed. These included: safety rod dynamic response in scram scenarios, flexibility and mishandling behavior, and response to thermal excursions associated with gamma heating. A surveillance program aimed at evaluating the integrity of the safety rods following actual operating conditions and justifying life extension for the rods was also proposed. Based on the experimental testing and analyses associated with this task, it is concluded that the boron carbide safety rod design meets the materials and mechanical criteria for successful operational performance.