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Sample records for nuclear safety asn

  1. [Implementing new technology in radiation oncology: The French agency for nuclear safety (ASN) report].

    PubMed

    Lartigau, É-F; Lisbona, A; Isambert, A; Cadot, P; Derreumaux, S; Dupuis, O; Gérard, J-P; Ledu, D; Mahé, M-A; Marchesi, V; Mazurier, J; De Oliveira, A; Phare, O; Aubert, B

    2015-10-01

    In August 2013, the French nuclear safety agency (ASN) requested the permanent group of experts in radiation protection in medicine (GPMED) to propose recommendations on the implementation of new technology and techniques in radiation oncology. These recommendations were finalized in February 2015 by the GPMED. In April 2015, the ASN sent a letter to the French ministry of health (DGS/DGOS), and its national health agencies (ANSM, INCa, HAS). In these letters, ASN proposed that, from the 12 recommendations made by the GPMED, an action plan should be established, whose control could be assigned to the French national cancer institute (INCa), as a pilot of the national committee for radiotherapy and that this proposal has to be considered at the next meeting of the national committee of radiotherapy. Copyright © 2015 Société française de radiothérapie oncologique (SFRO). Published by Elsevier SAS. All rights reserved.

  2. Nuclear safety

    NASA Technical Reports Server (NTRS)

    Buden, D.

    1991-01-01

    Topics dealing with nuclear safety are addressed which include the following: general safety requirements; safety design requirements; terrestrial safety; SP-100 Flight System key safety requirements; potential mission accidents and hazards; key safety features; ground operations; launch operations; flight operations; disposal; safety concerns; licensing; the nuclear engine for rocket vehicle application (NERVA) design philosophy; the NERVA flight safety program; and the NERVA safety plan.

  3. Nuclear Safety

    SciTech Connect

    Silver, E G

    1989-01-01

    This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  4. Nuclear criticality safety guide

    SciTech Connect

    Pruvost, N.L.; Paxton, H.C.

    1996-09-01

    This technical reference document cites information related to nuclear criticality safety principles, experience, and practice. The document also provides general guidance for criticality safety personnel and regulators.

  5. Revitalizing Nuclear Safety Research.

    ERIC Educational Resources Information Center

    National Academy of Sciences - National Research Council, Washington, DC.

    This report covers the general issues involved in nuclear safety research and points out the areas needing detailed consideration. Topics included are: (1) "Principles of Nuclear Safety Research" (examining who should fund, who should conduct, and who should set the agenda for nuclear safety research); (2) "Elements of a Future…

  6. Revitalizing Nuclear Safety Research.

    ERIC Educational Resources Information Center

    National Academy of Sciences - National Research Council, Washington, DC.

    This report covers the general issues involved in nuclear safety research and points out the areas needing detailed consideration. Topics included are: (1) "Principles of Nuclear Safety Research" (examining who should fund, who should conduct, and who should set the agenda for nuclear safety research); (2) "Elements of a Future…

  7. Nuclear Reactor Safety

    SciTech Connect

    Bales, J.D.; Boshears, R.

    1996-02-01

    Nuclear Reactor Safety (NRS), published monthly, is a collection of abstracts of worldwide information available on all safety-related aspects of reactors, including accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are other U.S. information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange, the International Atomic Energy Agency`s International Nuclear Information System, or government-to-government agreements. The digests in NRS and other citations to information on nuclear reactor safety dating from 1948 are available for online searching and retrieval in the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  8. Nuclear explosive safety study process

    SciTech Connect

    1997-01-01

    Nuclear explosives by their design and intended use require collocation of high explosives and fissile material. The design agencies are responsible for designing safety into the nuclear explosive and processes involving the nuclear explosive. The methodology for ensuring safety consists of independent review processes that include the national laboratories, Operations Offices, Headquarters, and responsible Area Offices and operating contractors with expertise in nuclear explosive safety. A NES Study is an evaluation of the adequacy of positive measures to minimize the possibility of an inadvertent or deliberate unauthorized nuclear detonation, high explosive detonation or deflagration, fire, or fissile material dispersal from the pit. The Nuclear Explosive Safety Study Group (NESSG) evaluates nuclear explosive operations against the Nuclear Explosive Safety Standards specified in DOE O 452.2 using systematic evaluation techniques. These Safety Standards must be satisfied for nuclear explosive operations.

  9. 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.

  10. Prospects for nuclear safety research

    SciTech Connect

    Beckjord, E.S.

    1995-04-01

    This document is the text of a paper presented by Eric S. Beckjord (Director, Nuclear Regulatory Research/NRC) at the 22nd Water Reactor Safety Meeting in Bethesda, MD in October 1994. The following topics are briefly reviewed: (1) Reactor vessel research, (2) Probabilistic risk assessment, (3) Direct containment heating, (4) Advanced LWR research, (5) Nuclear energy prospects in the US, and (6) Future nuclear safety research. Subtopics within the last category include economics, waste disposal, and health and safety.

  11. 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.

  12. Nuclear Powerplant Safety: Operations.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Nuclear Energy Office.

    Powerplant systems and procedures that ensure the day-to-day health and safety of people in and around the plant is referred to as operational safety. This safety is the result of careful planning, good engineering and design, strict licensing and regulation, and environmental monitoring. Procedures that assure operational safety at nuclear…

  13. Nuclear power: Siting and safety

    SciTech Connect

    Openshaw, S.

    1986-01-01

    By 2030, half, or even two-thirds, of all electricity may be generated by nuclear power. Major reactor accidents are still expected to be rare occurrences, but nuclear safety is largely a matter of faith. Terrorist attacks, sabotage, and human error could cause a significant accident. Reactor siting can offer an additional, design-independent margin of safety. Remote geographical sites for new plants would minimize health risks, protect the industry from negative changes in public opinion concerning nuclear energy, and improve long-term public acceptance of nuclear power. U.K. siting practices usually do not consider the contribution to safety that could be obtained from remote sites. This book discusses the present trends of siting policies of nuclear power and their design-independent margin of safety.

  14. NRC - regulator of nuclear safety

    SciTech Connect

    1997-05-01

    The U.S. Nuclear Regulatory Commission (NRC) was formed in 1975 to regulate the various commercial and institutional uses of nuclear energy, including nuclear power plants. The agency succeeded the Atomic Energy Commission, which previously had responsibility for both developing and regulating nuclear activities. Federal research and development work for all energy sources, as well as nuclear weapons production, is now conducted by the U.S. Department of Energy. Under its responsibility to protect public health and safety, the NRC has three principal regulatory functions: (1) establish standards and regulations, (2) issue licenses for nuclear facilities and users of nuclear materials, and (3) inspect facilities and users of nuclear materials to ensure compliance with the requirements. These regulatory functions relate to both nuclear power plants and to other uses of nuclear materials - like nuclear medicine programs at hospitals, academic activities at educational institutions, research work, and such industrial applications as gauges and testing equipment. The NRC places a high priority on keeping the public informed of its work. The agency recognizes the interest of citizens in what it does through such activities as maintaining public document rooms across the country and holding public hearings, public meetings in local areas, and discussions with individuals and organizations.

  15. Nuclear Safety for Space Systems

    NASA Astrophysics Data System (ADS)

    Offiong, Etim

    2010-09-01

    It is trite, albeit a truism, to say that nuclear power can provide propulsion thrust needed to launch space vehicles and also, to provide electricity for powering on-board systems, especially for missions to the Moon, Mars and other deep space missions. Nuclear Power Sources(NPSs) are known to provide more capabilities than solar power, fuel cells and conventional chemical means. The worry has always been that of safety. The earliest superpowers(US and former Soviet Union) have designed and launched several nuclear-powered systems, with some failures. Nuclear failures and accidents, however little the number, could be far-reaching geographically, and are catastrophic to humans and the environment. Building on the numerous research works on nuclear power on Earth and in space, this paper seeks to bring to bear, issues relating to safety of space systems - spacecrafts, astronauts, Earth environment and extra terrestrial habitats - in the use and application of nuclear power sources. It also introduces a new formal training course in Space Systems Safety.

  16. Autoclave nuclear criticality safety analysis

    SciTech Connect

    D`Aquila, D.M.; Tayloe, R.W. Jr.

    1991-12-31

    Steam-heated autoclaves are used in gaseous diffusion uranium enrichment plants to heat large cylinders of UF{sub 6}. Nuclear criticality safety for these autoclaves is evaluated. To enhance criticality safety, systems are incorporated into the design of autoclaves to limit the amount of water present. These safety systems also increase the likelihood that any UF{sub 6} inadvertently released from a cylinder into an autoclave is not released to the environment. Up to 140 pounds of water can be held up in large autoclaves. This mass of water is sufficient to support a nuclear criticality when optimally combined with 125 pounds of UF{sub 6} enriched to 5 percent U{sup 235}. However, water in autoclaves is widely dispersed as condensed droplets and vapor, and is extremely unlikely to form a critical configuration with released UF{sub 6}.

  17. Nuclear Criticality Safety Data Book

    SciTech Connect

    Hollenbach, D. F.

    2016-11-14

    The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

  18. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS ENVIRONMENT, CONSERVATION, OCCUPATIONAL SAFETY, AND DRUG-FREE WORKPLACE Environmental, Energy and Water Efficiency, Renewable Energy Technologies, and Occupational Safety Programs 923.7001 Nuclear safety. The...

  19. Safe use of atomic (Nuclear) power (Nuclear Safety)

    NASA Astrophysics Data System (ADS)

    Sidorenko, V. A.

    2013-12-01

    The established concept of ensuring safety for nuclear power sources is presented; the influence of severe accidents on nuclear power development is considered, including the accident at a Japan NPP in 2011, as well as the role of state regulation of nuclear safety.

  20. Elements of a nuclear criticality safety program

    SciTech Connect

    Hopper, C.M.

    1995-07-01

    Nuclear criticality safety programs throughout the United States are quite successful, as compared with other safety disciplines, at protecting life and property, especially when regarded as a developing safety function with no historical perspective for the cause and effect of process nuclear criticality accidents before 1943. The programs evolved through self-imposed and regulatory-imposed incentives. They are the products of conscientious individuals, supportive corporations, obliged regulators, and intervenors (political, public, and private). The maturing of nuclear criticality safety programs throughout the United States has been spasmodic, with stability provided by the volunteer standards efforts within the American Nuclear Society. This presentation provides the status, relative to current needs, for nuclear criticality safety program elements that address organization of and assignments for nuclear criticality safety program responsibilities; personnel qualifications; and analytical capabilities for the technical definition of critical, subcritical, safety and operating limits, and program quality assurance.

  1. Nuclear criticality safety: 2-day training course

    SciTech Connect

    Schlesser, J.A.

    1997-02-01

    This compilation of notes is presented as a source reference for the criticality safety course. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used as Los Alamos; be able to identify examples of circumstances present during criticality accidents; have participated in conducting two critical experiments; be asked to complete a critique of the nuclear criticality safety training course.

  2. A philosophy for space nuclear systems safety

    SciTech Connect

    Marshall, A.C.

    1992-08-01

    The unique requirements and contraints of space nuclear systems require careful consideration in the development of a safety policy. The Nuclear Safety Policy Working Group (NSPWG) for the Space Exploration Initiative has proposed a hierarchical approach with safety policy at the top of the hierarchy. This policy allows safety requirements to be tailored to specific applications while still providing reassurance to regulators and the general public that the necessary measures have been taken to assure safe application of space nuclear systems. The safety policy used by the NSPWG is recommended for all space nuclear programs and missions.

  3. Radiation Safety in Nuclear Medicine Procedures.

    PubMed

    Cho, Sang-Geon; Kim, Jahae; Song, Ho-Chun

    2017-03-01

    Since the nuclear disaster at the Fukushima Daiichi Nuclear Power Plant in 2011, radiation safety has become an important issue in nuclear medicine. Many structured guidelines or recommendations of various academic societies or international campaigns demonstrate important issues of radiation safety in nuclear medicine procedures. There are ongoing efforts to fulfill the basic principles of radiation protection in daily nuclear medicine practice. This article reviews important principles of radiation protection in nuclear medicine procedures. Useful references, important issues, future perspectives of the optimization of nuclear medicine procedures, and diagnostic reference level are also discussed.

  4. Nuclear criticality safety: 5-day training course

    SciTech Connect

    Schlesser, J.A.

    1992-11-01

    This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course's primary instructor. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used at Los Alamos; be able to identify examples of circumstances present during criticality accidents; be able to identify examples of computer codes used by the nuclear criticality safety specialist; be able to identify examples of safety consciousness required in nuclear criticality safety.

  5. Nuclear criticality safety: 5-day training course

    SciTech Connect

    Schlesser, J.A.

    1992-11-01

    This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course`s primary instructor. At the completion of this training course, the attendee will: be able to define terms commonly used in nuclear criticality safety; be able to appreciate the fundamentals of nuclear criticality safety; be able to identify factors which affect nuclear criticality safety; be able to identify examples of criticality controls as used at Los Alamos; be able to identify examples of circumstances present during criticality accidents; be able to identify examples of computer codes used by the nuclear criticality safety specialist; be able to identify examples of safety consciousness required in nuclear criticality safety.

  6. Safety culture in the nuclear versus non-nuclear organization

    SciTech Connect

    Haber, S.B.; Shurberg, D.A.

    1996-10-01

    The importance of safety culture in the safe and reliable operation of nuclear organizations is not a new concept. The greatest barriers to this area of research are twofold: (1) the definition and criteria of safety culture for a nuclear organization and (2) the measurement of those attributes in an objective and systematic fashion. This paper will discuss a proposed resolution of those barriers as demonstrated by the collection of data across nuclear and non-nuclear facilities over a two year period.

  7. Software Quality Assurance for Nuclear Safety Systems

    SciTech Connect

    Sparkman, D R; Lagdon, R

    2004-05-16

    The US Department of Energy has undertaken an initiative to improve the quality of software used to design and operate their nuclear facilities across the United States. One aspect of this initiative is to revise or create new directives and guides associated with quality practices for the safety software in its nuclear facilities. Safety software includes the safety structures, systems, and components software and firmware, support software and design and analysis software used to ensure the safety of the facility. DOE nuclear facilities are unique when compared to commercial nuclear or other industrial activities in terms of the types and quantities of hazards that must be controlled to protect workers, public and the environment. Because of these differences, DOE must develop an approach to software quality assurance that ensures appropriate risk mitigation by developing a framework of requirements that accomplishes the following goals: {sm_bullet} Ensures the software processes developed to address nuclear safety in design, operation, construction and maintenance of its facilities are safe {sm_bullet} Considers the larger system that uses the software and its impacts {sm_bullet} Ensures that the software failures do not create unsafe conditions Software designers for nuclear systems and processes must reduce risks in software applications by incorporating processes that recognize, detect, and mitigate software failure in safety related systems. It must also ensure that fail safe modes and component testing are incorporated into software design. For nuclear facilities, the consideration of risk is not necessarily sufficient to ensure safety. Systematic evaluation, independent verification and system safety analysis must be considered for software design, implementation, and operation. The software industry primarily uses risk analysis to determine the appropriate level of rigor applied to software practices. This risk-based approach distinguishes safety

  8. ASN reputation system model

    NASA Astrophysics Data System (ADS)

    Hutchinson, Steve; Erbacher, Robert F.

    2015-05-01

    Network security monitoring is currently challenged by its reliance on human analysts and the inability for tools to generate indications and warnings for previously unknown attacks. We propose a reputation system based on IP address set membership within the Autonomous System Number (ASN) system. Essentially, a metric generated based on the historic behavior, or misbehavior, of nodes within a given ASN can be used to predict future behavior and provide a mechanism to locate network activity requiring inspection. This will provide reinforcement of notifications and warnings and lead to inspection for ASNs known to be problematic even if initial inspection leads to interpretation of the event as innocuous. We developed proof of concept capabilities to generate the IP address to ASN set membership and analyze the impact of the results. These results clearly show that while some ASNs are one-offs with individual or small numbers of misbehaving IP addresses, there are definitive ASNs with a history of long term and wide spread misbehaving IP addresses. These ASNs with long histories are what we are especially interested in and will provide an additional correlation metric for the human analyst and lead to new tools to aid remediation of these IP address blocks.

  9. The history of nuclear weapon safety devices

    SciTech Connect

    Plummer, D.W.; Greenwood, W.H.

    1998-06-01

    The paper presents the history of safety devices used in nuclear weapons from the early days of separables to the latest advancements in MicroElectroMechanical Systems (MEMS). Although the paper focuses on devices, the principles of Enhanced Nuclear Detonation Safety implementation will also be presented.

  10. Nuclear Powerplant Safety: Design and Planning.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Nuclear Energy Office.

    The most important concern in the design, construction and operation of nuclear powerplants is safety. Nuclear power is one of the major contributors to the nation's supply of electricity; therefore, it is important to assure its safe use. Each different type of powerplant has special design features and systems to protect health and safety. One…

  11. Nuclear criticality safety: 3-day training course

    SciTech Connect

    Schlesser, J.A.

    1992-11-01

    This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course`s primary instructor. At the completion of this training course, the attendee will: (1) be able to define terms commonly used in nuclear criticality safety; (2) be able to appreciate the fundamentals of nuclear criticality safety; (3) be able to identify factors which affect nuclear criticality safety; (4) be able to identify examples of criticality controls as used at Los Alamos; (5) be able to identify examples of circumstances present during criticality accidents; (6) be able to identify examples of safety consciousness required in nuclear criticality safety.

  12. Nuclear criticality safety: 3-day training course

    SciTech Connect

    Schlesser, J.A.

    1992-11-01

    This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course's primary instructor. At the completion of this training course, the attendee will: (1) be able to define terms commonly used in nuclear criticality safety; (2) be able to appreciate the fundamentals of nuclear criticality safety; (3) be able to identify factors which affect nuclear criticality safety; (4) be able to identify examples of criticality controls as used at Los Alamos; (5) be able to identify examples of circumstances present during criticality accidents; (6) be able to identify examples of safety consciousness required in nuclear criticality safety.

  13. Advanced research workshop: nuclear materials safety

    SciTech Connect

    Jardine, L J; Moshkov, M M

    1999-01-28

    The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on the storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of

  14. Manned space flight nuclear system safety. Volume 5: Nuclear System safety guidelines. Part 1: Space base nuclear safety

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The design and operations guidelines and requirements developed in the study of space base nuclear system safety are presented. Guidelines and requirements are presented for the space base subsystems, nuclear hardware (reactor, isotope sources, dynamic generator equipment), experiments, interfacing vehicles, ground support systems, range safety and facilities. Cross indices and references are provided which relate guidelines to each other, and to substantiating data in other volumes. The guidelines are intended for the implementation of nuclear safety related design and operational considerations in future space programs.

  15. Nuclear safety policy working group recommendations on nuclear propulsion safety for the space exploration initiative

    NASA Technical Reports Server (NTRS)

    Marshall, Albert C.; Lee, James H.; Mcculloch, William H.; Sawyer, J. Charles, Jr.; Bari, Robert A.; Cullingford, Hatice S.; Hardy, Alva C.; Niederauer, George F.; Remp, Kerry; Rice, John W.

    1993-01-01

    An interagency Nuclear Safety Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program. These recommendations, which are contained in this report, should facilitate the implementation of mission planning and conceptual design studies. The NSPWG has recommended a top-level policy to provide the guiding principles for the development and implementation of the SEI nuclear propulsion safety program. In addition, the NSPWG has reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. These recommendations should be useful for the development of the program's top-level requirements for safety functions (referred to as Safety Functional Requirements). The safety requirements and guidelines address the following topics: reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations.

  16. Nuclear Powerplant Safety: Source Terms. Nuclear Energy.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Nuclear Energy Office.

    There has been increased public interest in the potential effects of nuclear powerplant accidents since the Soviet reactor accident at Chernobyl. People have begun to look for more information about the amount of radioactivity that might be released into the environment as a result of such an accident. When this issue is discussed by people…

  17. Manned space flight nuclear system safety. Voluem 5: Nuclear system safety guidelines. Part 2: Space shuttle/nuclear payloads safety

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The design and operations guidelines and requirements developed in the study of space shuttle nuclear system transportation are presented. Guidelines and requirements are presented for the shuttle, nuclear payloads (reactor, isotope-Brayton and small isotope sources), ground support systems and facilities. Cross indices and references are provided which relate guidelines to each other, and to substantiating data in other volumes. The guidelines are intended for the implementation of nuclear safety related design and operational considerations in future space programs.

  18. Nuclear criticality safety department training implementation

    SciTech Connect

    Carroll, K.J.; Taylor, R.G.; Worley, C.A.

    1996-09-06

    The Nuclear Criticality Safety Department (NCSD) is committed to developing and maintaining a staff of qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. The NCSD Qualification Program is described in Y/DD-694, Qualification Program, Nuclear Criticality Safety Department This document provides a listing of the roles and responsibilities of NCSD personnel with respect to training and details of the Training Management System (TMS) programs, Mentoring Checklists and Checksheets, as well as other documentation utilized to implement the program. This document supersedes Y/DD-696, Revision 2, dated 3/27/96, Training Implementation, Nuclear Criticality Safety Department. There are no backfit requirements associated with revisions to this document.

  19. Comparison of radiation safety and nuclear explosive safety disciplines

    SciTech Connect

    Winstanley, J. L.

    1998-10-10

    In August 1945, U.S. Navy Captain William Parsons served as the weaponeer aboard the Enola Gay for the mission to Hiroshima (Shelton 1988). In view of the fact that four B-29s had crashed and burned on takeoff from Tinian the night before, Captain Parsons made the decision to arm the gun-type weapon after takeoff for safety reasons (15 kilotons of TNT equivalent). Although he had no control over the success of the takeoff, he could prevent the possibility of a nuclear detonation on Tinian by controlling what we now call the nuclear explosive. As head of the Ordnance Division at Los Alamos and a former gunnery officer, Captain Parsons clearly understood the role of safety in his work. The advent of the pre-assembled implosion weapon where the high explosive and nuclear materials are always in an intimate configuration meant that nuclear explosive safety became a reality at a certain point in development and production not just at the time of delivery by the military. This is the only industry where nuclear materials are intentionally put in contact with high explosives. The agency of the U.S. Government responsible for development and production of U.S. nuclear weapons is the Department of Energy (DOE) (and its predecessor agencies). This paper will be limited to nuclear explosive safety as it is currently practiced within the DOE nuclear weapons

  20. Nuclear data for criticality safety - current issues

    SciTech Connect

    Leal, L.C.; Jordan, W.C.; Wright, R.Q.

    1995-06-01

    Traditionally, nuclear data evaluations have been performed in support of the analysis and design of thermal and fast reactors. In general, the neutron spectra characteristic of the thermal and fast systems used for data testing are predominantly in the low- and high-energy range with a relatively small influence from the intermediate-energy range. In the area of nuclear criticality safety, nuclear systems arising from applications involving fissionable materials outside reactors can lead to situations very different from those most commonly found in reactor analysis and design. These systems are not limited to thermal or fast and may have significant influence from the intermediate energy range. The extension of the range of applicability of the nuclear data evaluation beyond thermal and fast systems is therefore needed to cover problems found in nuclear criticality safety. Before criticality safety calculations are performed, the bias and uncertainties of the codes and cross sections that are used must be determined. The most common sources of uncertainties, in general, are the calculational methodologies and the uncertainties related to the nuclear data, such as the microscopic cross sections, entering into the calculational procedure. The aim here is to focus on the evaluated nuclear data pertaining to applications in nuclear criticality safety.

  1. 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.

  2. Nuclear safety for the space exploration initiative

    NASA Technical Reports Server (NTRS)

    Dix, Terry E.

    1991-01-01

    The results of a study to identify potential hazards arising from nuclear reactor power systems for use on the lunar and Martian surfaces, related safety issues, and resolutions of such issues by system design changes, operating procedures, and other means are presented. All safety aspects of nuclear reactor power systems from prelaunch ground handling to eventual disposal were examined consistent with the level of detail for SP-100 reactor design at the 1988 System Design Review and for launch vehicle and space transport vehicle designs and mission descriptions as defined in the 90-day Space Exploration Initiative (SEI) study. Information from previous aerospace nuclear safety studies was used where appropriate. Safety requirements for the SP-100 space nuclear reactor system were compiled. Mission profiles were defined with emphasis on activities after low earth orbit insertion. Accident scenarios were then qualitatively defined for each mission phase. Safety issues were identified for all mission phases with the aid of simplified event trees. Safety issue resolution approaches of the SP-100 program were compiled. Resolution approaches for those safety issues not covered by the SP-100 program were identified. Additionally, the resolution approaches of the SP-100 program were examined in light of the moon and Mars missions.

  3. Nuclear Criticality Safety Application Guide: Safety Analysis Report Update Program

    SciTech Connect

    Not Available

    1994-02-01

    Martin Marietta Energy Systems, Inc. (MMES) is committed to performing and documenting safety analyses for facilities it manages for the Department of Energy (DOE). Safety analyses are performed to identify hazards and potential accidents; to analyze the adequacy of measures taken to eliminate, control, or mitigate hazards; and to evaluate potential accidents and determine associated risks. Safety Analysis Reports (SARs) are prepared to document the safety analysis to ensure facilities can be operated safely and in accordance with regulations. Many of the facilities requiring a SAR process fissionable material creating the potential for a nuclear criticality accident. MMES has long had a nuclear criticality safety program that provides the technical support to fissionable material operations to ensure the safe processing and storage of fissionable materials. The guiding philosophy of the program has always been the application of the double-contingency principle, which states: {open_quotes}process designs shall incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible.{close_quotes} At Energy Systems analyses have generally been maintained to document that no single normal or abnormal operating conditions that could reasonably be expected to occur can cause a nuclear criticality accident. This application guide provides a summary description of the MMES Nuclear Criticality Safety Program and the MMES Criticality Accident Alarm System requirements for inclusion in facility SARs. The guide also suggests a way to incorporate the analyses conducted pursuant to the double-contingency principle into the SAR. The prime objective is to minimize duplicative effort between the NCSA process and the SAR process and yet adequately describe the methodology utilized to prevent a nuclear criticality accident.

  4. Neutron nuclear data measurements for criticality safety

    NASA Astrophysics Data System (ADS)

    Guber, Klaus; Paradela, Carlos; Heyse, Jan; Kopecky, Stefan; Schillebeeckx, Peter; Siegler, Peter

    2017-09-01

    To support the US Department of Energy Nuclear Criticality Safety Program, neutron-induced cross section experiments were performed at the Geel Electron Linear Accelerator of the Joint Research Center Site Geel, European Union. Neutron capture and transmission measurements were carried out using metallic natural cerium and vanadium samples. Together with existing data, the measured data will be used for a new evaluation and will be submitted with covariances to the ENDF/B nuclear data library.

  5. Management of National Nuclear Power Programs for assured safety

    SciTech Connect

    Connolly, T.J.

    1985-01-01

    Topics discussed in this report include: nuclear utility organization; before the Florida Public Service Commission in re: St. Lucie Unit No. 2 cost recovery; nuclear reliability improvement and safety operations; nuclear utility management; training of nuclear facility personnel; US experience in key areas of nuclear safety; the US Nuclear Regulatory Commission - function and process; regulatory considerations of the risk of nuclear power plants; overview of the processes of reliability and risk management; management significance of risk analysis; international and domestic institutional issues for peaceful nuclear uses; the role of the Institute of Nuclear Power Operations (INPO); and nuclear safety activities of the International Atomic Energy Agency (IAEA).

  6. Nuclear safety technology and public acceptance

    NASA Astrophysics Data System (ADS)

    Kienle, F.

    1985-11-01

    In the years 1976 to 1982 officialdom intensified the safety regulations in German nuclear power plants out of all proportion, without actually bringing about a recognizable plus in safety or indeed a greater acceptance by the public of the peaceful use of nuclear energy. Although the risk to employees of nuclear power plants and to the population living in their vicinity is substantially smaller than the dangers of modern civilization, the general public still regards with concern the consequences of radioactive exposure and the hazards to later generations from long-life radionuclides. The task for the coming years must be to maintain the safety standard now attained, while simultaneously reducing those exaggerated individual requirements in order to establish a balance in safety precautions. Additionally, a proposal put forward by Sir Walter Marshall, Chairman of the CEGB, should be pursued, i.e., to put the presumed risks of nuclear energy into their correct perspective in the public eye using comprehensible comparisons such as the risks from active or passive smoking. This cannot be accomplished by quoting abstract statistics.

  7. TOPAZ-2 Nuclear Power System safety assurance

    SciTech Connect

    Nikitin, V.P.; Ogloblin, B.G.; Lutov, Y.I.; Luppov, A.N.; Shalaev, A.I. ); Ponomarev-Stepnoi, N.N.; Usov, V.A.; Nechaev, Y.A. )

    1993-01-15

    TOPAZ-2 Nuclear Power System (NPS) safety philosophy is based on the requirement that the reactor shall not be critical during all kinds of operations prior to its start-up on the safe orbit (except for physical start-up). Potentially dangerous operation were analyzed and both computational and experimental studies were carried out.

  8. A PIP chart for nuclear plant safety

    SciTech Connect

    Suzuki, Tatsujiro; Yamaoka, Taiji )

    1992-01-01

    While it is known that social and political aspects of nuclear safety issues are important, little study has been done on identifying the breadth of stakeholders whose policies have important influences over nuclear plant safety in a comprehensive way. The objectives of this study are to develop a chart that visually identifies important stakeholders and their policies and illustrates these influences in a hierarchical representation so that the relationship between stakeholders and nuclear safety will be better understood. This study is based on a series of extensive interviews with major stakeholders, such as nuclear plant managers, corporate planning vice presidents, state regulators, news media, and public interest groups, and focuses on one US nuclear power plant. Based on the interview results, the authors developed a conceptual policy influence paths (PIP) chart. The PIP chart illustrates the hierarchy of influence among stakeholders. The PIP chart is also useful in identifying possible stakeholders who can be easily overlooked without the PIP chart. In addition, it shows that influence flow is circular rather than linear in one direction.

  9. NUCLEAR SAFETY DESIGN BASES FOR LICENSE APPLICATION

    SciTech Connect

    R.J. Garrett

    2005-03-08

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111 [DIRS 156605] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113 [DIRS 156605] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period.

  10. Nuclear Safety Design Base for License Application

    SciTech Connect

    R.J. Garrett

    2005-09-29

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111(b) [DIRS 173273] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113(b) and (c) [DIRS 173273] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period.

  11. Nuclear safety research collaborations between the U.S. and Russian Federation International Nuclear Safety Centers

    SciTech Connect

    Hill, D. J.; Braun, J. C.; Klickman, A. E.; Bougaenko, S. E.; Kabonov, L. P.; Kraev, A. G.

    2000-05-05

    The Russian Federation Ministry for Atomic Energy (MINATOM) and the US Department of Energy (USDOE) have formed International Nuclear Safety Centers to collaborate on nuclear safety research. USDOE established the US Center (ISINSC) at Argonne National Laboratory (ANL) in October 1995. MINATOM established the Russian Center (RINSC) at the Research and Development Institute of Power Engineering (RDIPE) in Moscow in July 1996. In April 1998 the Russian center became a semi-independent, autonomous organization under MINATOM. The goals of the center are to: Cooperate in the development of technologies associated with nuclear safety in nuclear power engineering; Be international centers for the collection of information important for safety and technical improvements in nuclear power engineering; and Maintain a base for fundamental knowledge needed to design nuclear reactors. The strategic approach is being used to accomplish these goals is for the two centers to work together to use the resources and the talents of the scientists associated with the US Center and the Russian Center to do collaborative research to improve the safety of Russian-designed nuclear reactors. The two centers started conducting joint research and development projects in January 1997. Since that time the following ten joint projects have been initiated: INSC databases--web server and computing center; Coupled codes--Neutronic and thermal-hydraulic; Severe accident management for Soviet-designed reactors; Transient management and advanced control; Survey of relevant nuclear safety research facilities in the Russian Federation; Computer code validation for transient analysis of VVER and RBMK reactors; Advanced structural analysis; Development of a nuclear safety research and development plan for MINATOM; Properties and applications of heavy liquid metal coolants; and Material properties measurement and assessment. Currently, there is activity in eight of these projects. Details on each of these

  12. Space nuclear safety from a user's viewpoint

    NASA Technical Reports Server (NTRS)

    Campbell, R. W.

    1985-01-01

    The National Aeronautics and Space Administration (NASA) launched the Jet Propulsion Laboratory's (JPL) two Voyager spacecraft to Jupiter in 1977, each using three radioisotope thermoelectric generators (RTGs) supplied by the Department of Energy (DOE) for onboard electric power. In 1986 NASA will launch JPL's Galileo spacecraft to Jupiter equipped with two DOE supplied RTGs of an improved design. NASA and JPL are also responsible for obtaining a single RTG of this type from DOE and supplying it to the European Space Agency as part of its participation in the International Solar Polar Mission. As a result of these missions, JPL has been deeply involved in space nuclear safety as a user. This paper will give a brief review of the user contributions by JPL - and NASA in general - to the nuclear safety processes and relate them to the overall nuclear safety program necessary for the launch of an RTG. The two major safety areas requiring user support are the ground operations involving RTGs at the launch site and the failure modes and probabilities associated with launch accidents.

  13. Nuclear Safety Information Center, Its Products and Services

    ERIC Educational Resources Information Center

    Buchanan, J. R.

    1970-01-01

    The Nuclear Safety Information Center (NSIC) serves as a focal point for the collection, analysis and dissemination of information related to safety problems encountered in the design, analysis, and operation of nuclear facilities. (Author/AB)

  14. Nuclear Safety Information Center, Its Products and Services

    ERIC Educational Resources Information Center

    Buchanan, J. R.

    1970-01-01

    The Nuclear Safety Information Center (NSIC) serves as a focal point for the collection, analysis and dissemination of information related to safety problems encountered in the design, analysis, and operation of nuclear facilities. (Author/AB)

  15. Safety of Decommissioning of Nuclear Facilities

    SciTech Connect

    Batandjieva, B.; Warnecke, E.; Coates, R.

    2008-01-15

    Full text of publication follows: ensuring safety during all stages of facility life cycle is a widely recognised responsibility of the operators, implemented under the supervision of the regulatory body and other competent authorities. As the majority of the facilities worldwide are still in operation or shutdown, there is no substantial experience in decommissioning and evaluation of safety during decommissioning in majority of Member States. The need for cooperation and exchange of experience and good practices on ensuring and evaluating safety of decommissioning was one of the outcomes of the Berlin conference in 2002. On this basis during the last three years IAEA initiated a number of international projects that can assist countries, in particular small countries with limited resources. The main IAEA international projects addressing safety during decommissioning are: (i) DeSa Project on Evaluation and Demonstration of Safety during Decommissioning; (ii) R{sup 2}D{sup 2}P project on Research Reactors Decommissioning Demonstration Project; and (iii) Project on Evaluation and Decommissioning of Former Facilities that used Radioactive Material in Iraq. This paper focuses on the DeSa Project activities on (i) development of a harmonised methodology for safety assessment for decommissioning; (ii) development of a procedure for review of safety assessments; (iii) development of recommendations on application of the graded approach to the performance and review of safety assessments; and (iv) application of the methodology and procedure to the selected real facilities with different complexities and hazard potentials (a nuclear power plant, a research reactor and a nuclear laboratory). The paper also outlines the DeSa Project outcomes and planned follow-up activities. It also summarises the main objectives and activities of the Iraq Project and introduces the R{sup 2}D{sup 2} Project, which is a subject of a complementary paper.

  16. The Interagency Nuclear Safety Review Panel's Galileo safety evaluation report

    SciTech Connect

    Nelson, R.C.; Gray, L.B.; Huff, D.A.

    1989-01-01

    The safety evaluation report (SER) for Galileo was prepared by the Interagency Nuclear Safety Review Panel (INSRP) coordinators in accordance with Presidential directive/National Security Council memorandum 25. The INSRP consists of three coordinators appointed by their respective agencies, the Department of Defense, the Department of Energy (DOE), and the National Aeronautics and Space Administration (NASA). These individuals are independent of the program being evaluated and depend on independent experts drawn from the national technical community to serve on the five INSRP subpanels. The Galileo SER is based on input provided by the NASA Galileo Program Office, review and assessment of the final safety analysis report prepared by the Office of Special Applications of the DOE under a memorandum of understanding between NASA and the DOE, as well as other related data and analyses. The SER was prepared for use by the agencies and the Office of Science and Technology Policy, Executive Office of the Present for use in their launch decision-making process. Although more than 20 nuclear-powered space missions have been previously reviewed via the INSRP process, the Galileo review constituted the first review of a nuclear power source associated with launch aboard the Space Transportation System.

  17. 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)

  18. 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)

  19. Double-clad nuclear fuel safety rod

    DOEpatents

    McCarthy, William H.; Atcheson, Donald B.; Vaidyanathan, Swaminathan

    1984-01-01

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  20. New Improved Nuclear Data for Nuclear Criticality and Safety

    SciTech Connect

    Guber, Klaus H; Leal, Luiz C; Lampoudis, C.; Kopecky, S.; Schillebeeckx, P.; Emiliani, F.; Wynants, R.; Siegler, P.

    2011-01-01

    The Geel Electron Linear Accelerator (GELINA) was used to measure neutron total and capture cross sections of {sup 182,183,184,186}W and {sup 63,65}Cu in the energy range from 100 eV to {approx}200 keV using the time-of-flight method. GELINA is the only high-power white neutron source with excellent timing resolution and ideally suited for these experiments. Concerns about the use of existing cross-section data in nuclear criticality calculations using Monte Carlo codes and benchmarks were a prime motivator for the new cross-section measurements. To support the Nuclear Criticality Safety Program, neutron cross-section measurements were initiated using GELINA at the EC-JRC-IRMM. Concerns about data deficiencies in some existing cross-section evaluations from libraries such as ENDF/B, JEFF, or JENDL for nuclear criticality calculations were the prime motivator for new cross-section measurements. Over the past years many troubles with existing nuclear data have emerged, such as problems related to proper normalization, neutron sensitivity backgrounds, poorly characterized samples, and use of improper pulse-height weighting functions. These deficiencies may occur in the resolved- and unresolved-resonance region and may lead to erroneous nuclear criticality calculations. An example is the use of the evaluated neutron cross-section data for tungsten in nuclear criticality safety calculations, which exhibit discrepancies in benchmark calculations and show the need for reliable covariance data. We measured the neutron total and capture cross sections of {sup 182,183,184,186}W and {sup 63,65}Cu in the neutron energy range from 100 eV to several hundred keV. This will help to improve the representation of the cross sections since most of the available evaluated data rely only on old measurements. Usually these measurements were done with poor experimental resolution or only over a very limited energy range, which is insufficient for the current application.

  1. Manned space flight nuclear system safety. Volume 6: Space base nuclear system safety plan

    NASA Technical Reports Server (NTRS)

    1972-01-01

    A qualitative identification of the steps required to assure the incorporation of radiological system safety principles and objectives into all phases of a manned space base program are presented. Specific areas of emphasis include: (1) radiological program management, (2) nuclear system safety plan implementation, (3) impact on program, and (4) summary of the key operation and design guidelines and requirements. The plan clearly indicates the necessity of considering and implementing radiological system safety recommendations as early as possible in the development cycle to assure maximum safety and minimize the impact on design and mission plans.

  2. Nuclear Safety. Technical progress journal: Volume 35, No.2

    SciTech Connect

    1994-09-01

    This journal covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and de commissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

  3. Nuclear Safety: Technical progress review, January-March 1988

    SciTech Connect

    Silver, E G

    1988-01-01

    This journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  4. Nuclear Safety: Technical progress review, January--March 1989

    SciTech Connect

    Silver, E. G.

    1989-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  5. Safety in nuclear power plants in India.

    PubMed

    Deolalikar, R

    2008-12-01

    Safety in nuclear power plants (NPPs) in India is a very important topic and it is necessary to dissipate correct information to all the readers and the public at large. In this article, I have briefly described how the safety in our NPPs is maintained. Safety is accorded overriding priority in all the activities. NPPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, dosimetry, approved standard operating and maintenance procedures, a well-defined waste management methodology, proper well documented and periodically rehearsed emergency preparedness and disaster management plans. The NPPs have occupational health policies covering periodic medical examinations, dosimetry and bioassay and are backed-up by fully equipped Personnel Decontamination Centers manned by doctors qualified in Occupational and Industrial Health. All the operating plants are ISO 14001 and IS 18001 certified plants. The Nuclear Power Corporation of India Limited today has 17 operating plants and five plants under construction, and our scientists and engineers are fully geared to take up many more in order to meet the national requirements.

  6. Safety in nuclear power plants in India

    PubMed Central

    Deolalikar, R.

    2008-01-01

    Safety in nuclear power plants (NPPs) in India is a very important topic and it is necessary to dissipate correct information to all the readers and the public at large. In this article, I have briefly described how the safety in our NPPs is maintained. Safety is accorded overriding priority in all the activities. NPPs in India are not only safe but are also well regulated, have proper radiological protection of workers and the public, regular surveillance, dosimetry, approved standard operating and maintenance procedures, a well-defined waste management methodology, proper well documented and periodically rehearsed emergency preparedness and disaster management plans. The NPPs have occupational health policies covering periodic medical examinations, dosimetry and bioassay and are backed-up by fully equipped Personnel Decontamination Centers manned by doctors qualified in Occupational and Industrial Health. All the operating plants are ISO 14001 and IS 18001 certified plants. The Nuclear Power Corporation of India Limited today has 17 operating plants and five plants under construction, and our scientists and engineers are fully geared to take up many more in order to meet the national requirements. PMID:20040970

  7. Nuclear Safety: Volume 29, No. 3: Technical progress review

    SciTech Connect

    Silver, E G

    1988-07-01

    Nuclear Safety is a review journal that covers significant development in the field of nuclear safety. Its scope included the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Individual papers have been cataloged separately.

  8. HANFORD NUCLEAR CRITICALITY SAFETY PROGRAM DATABASE

    SciTech Connect

    TOFFER, H.

    2005-05-02

    The Hanford Database is a useful information retrieval tool for a criticality safety practitioner. The database contains nuclear criticality literature screened for parameter studies. The entries, characterized with a value index, are segregated into 16 major and six minor categories. A majority of the screened entries have abstracts and a limited number are connected to the Office of Scientific and Technology Information (OSTI) database of full-size documents. Simple and complex searches of the data can be accomplished very rapidly and the end-product of the searches could be a full-size document. The paper contains a description of the database, user instructions, and a number of examples.

  9. Manned space flight nuclear system safety. Volume 4: Space shuttle nuclear system transportation. Part 1: Space shuttle nuclear safety

    NASA Technical Reports Server (NTRS)

    1972-01-01

    An analysis of the nuclear safety aspects (design and operational considerations) in the transport of nuclear payloads to and from earth orbit by the space shuttle is presented. Three representative nuclear payloads used in the study were: (1) the zirconium hydride reactor Brayton power module, (2) the large isotope Brayton power system and (3) small isotopic heat sources which can be a part of an upper stage or part of a logistics module. Reference data on the space shuttle and nuclear payloads are presented in an appendix. Safety oriented design and operational requirements were identified to integrate the nuclear payloads in the shuttle mission. Contingency situations were discussed and operations and design features were recommended to minimize the nuclear hazards. The study indicates the safety, design and operational advantages in the use of a nuclear payload transfer module. The transfer module can provide many of the safety related support functions (blast and fragmentation protection, environmental control, payload ejection) minimizing the direct impact on the shuttle.

  10. Information Services at the Nuclear Safety Analysis Center.

    ERIC Educational Resources Information Center

    Simard, Ronald

    This paper describes the operations of the Nuclear Safety Analysis Center. Established soon after an accident at the Three Mile Island nuclear power plant near Harrisburg, Pennsylvania, its efforts were initially directed towards a detailed analysis of the accident. Continuing functions include: (1) the analysis of generic nuclear safety issues,…

  11. Information Services at the Nuclear Safety Analysis Center.

    ERIC Educational Resources Information Center

    Simard, Ronald

    This paper describes the operations of the Nuclear Safety Analysis Center. Established soon after an accident at the Three Mile Island nuclear power plant near Harrisburg, Pennsylvania, its efforts were initially directed towards a detailed analysis of the accident. Continuing functions include: (1) the analysis of generic nuclear safety issues,…

  12. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 10 Energy 2 2011-01-01 2011-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.124 Criteria for nuclear criticality safety. (a) Design for...

  13. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... 10 Energy 2 2013-01-01 2013-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.124 Criteria for nuclear criticality safety. (a) Design for...

  14. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... 10 Energy 2 2012-01-01 2012-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.124 Criteria for nuclear criticality safety. (a) Design for...

  15. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... 10 Energy 2 2014-01-01 2014-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.124 Criteria for nuclear criticality safety. (a) Design for...

  16. 10 CFR 72.124 - Criteria for nuclear criticality safety.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Criteria for nuclear criticality safety. 72.124 Section 72... STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE General Design Criteria § 72.124 Criteria for nuclear criticality safety. (a) Design for...

  17. Applicability of trends in nuclear safety analysis to space nuclear power systems

    SciTech Connect

    Bari, R.A.

    1992-10-01

    A survey is presented of some current trends in nuclear safety analysis that may be relevant to space nuclear power systems. This includes: lessons learned from operating power reactor safety and licensing; approaches to the safety design of advanced and novel reactors and facilities; the roles of risk assessment, extremely unlikely accidents, safety goals/targets; and risk-benefit analysis and communication.

  18. Manned space flight nuclear system safety. Volume 1: base nuclear system safety

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The mission and terrestrial nuclear safety aspects of future long duration manned space missions in low earth orbit are discussed. Nuclear hazards of a typical low earth orbit Space Base mission (from natural sources and on-board nuclear hardware) have been identified and evaluated. Some of the principal nuclear safety design and procedural considerations involved in launch, orbital, and end of mission operations are presented. Areas of investigation include radiation interactions with the crew, subsystems, facilities, experiments, film, interfacing vehicles, nuclear hardware and the terrestrial populace. Results of the analysis indicate: (1) the natural space environment can be the dominant radiation source in a low earth orbit where reactors are effectively shielded, (2) with implementation of safety guidelines the reactor can present a low risk to the crew, support personnel, the terrestrial populace, flight hardware and the mission, (3) ten year missions are feasible without exceeding integrated radiation limits assigned to flight hardware, and (4) crew stay-times up to one year are feasible without storm shelter provisions.

  19. Tutorial on nuclear thermal propulsion safety for Mars

    SciTech Connect

    Buden, D.

    1992-01-01

    Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments.

  20. Tutorial on nuclear thermal propulsion safety for Mars

    SciTech Connect

    Buden, D.

    1992-08-01

    Safety is the prime design requirement for nuclear thermal propulsion (NTP). It must be built in at the initiation of the design process. An understanding of safety concerns is fundamental to the development of nuclear rockets for manned missions to Mars and many other applications that will be enabled or greatly enhanced by the use of nuclear propulsion. To provide an understanding of the basic issues, a tutorial has been prepared. This tutorial covers a range of topics including safety requirements and approaches to meet these requirements, risk and safety analysis methodology, NERVA reliability and safety approach, and life cycle risk assessments.

  1. Accurate Fission Data for Nuclear Safety

    NASA Astrophysics Data System (ADS)

    Solders, A.; Gorelov, D.; Jokinen, A.; Kolhinen, V. S.; Lantz, M.; Mattera, A.; Penttilä, H.; Pomp, S.; Rakopoulos, V.; Rinta-Antila, S.

    2014-05-01

    The Accurate fission data for nuclear safety (AlFONS) project aims at high precision measurements of fission yields, using the renewed IGISOL mass separator facility in combination with a new high current light ion cyclotron at the University of Jyväskylä. The 30 MeV proton beam will be used to create fast and thermal neutron spectra for the study of neutron induced fission yields. Thanks to a series of mass separating elements, culminating with the JYFLTRAP Penning trap, it is possible to achieve a mass resolving power in the order of a few hundred thousands. In this paper we present the experimental setup and the design of a neutron converter target for IGISOL. The goal is to have a flexible design. For studies of exotic nuclei far from stability a high neutron flux (1012 neutrons/s) at energies 1 - 30 MeV is desired while for reactor applications neutron spectra that resembles those of thermal and fast nuclear reactors are preferred. It is also desirable to be able to produce (semi-)monoenergetic neutrons for benchmarking and to study the energy dependence of fission yields. The scientific program is extensive and is planed to start in 2013 with a measurement of isomeric yield ratios of proton induced fission in uranium. This will be followed by studies of independent yields of thermal and fast neutron induced fission of various actinides.

  2. Requesting and granting exemptions to nuclear safety rules

    SciTech Connect

    Not Available

    1995-02-01

    This standard provides an acceptable process for requesting and granting exemptions to DOE Nuclear Safety rules. The provisions of 10 CFR Part 820.63 allow DOE to determine the procedures to be used to comply with the responsibilities regarding exemption relief from DOE nuclear safety rules. The procedure in this standard defines an acceptable method for meeting these responsibilities.

  3. Providing Nuclear Criticality Safety Analysis Education through Benchmark Experiment Evaluation

    SciTech Connect

    John D. Bess; J. Blair Briggs; David W. Nigg

    2009-11-01

    One of the challenges that today's new workforce of nuclear criticality safety engineers face is the opportunity to provide assessment of nuclear systems and establish safety guidelines without having received significant experience or hands-on training prior to graduation. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and/or the International Reactor Physics Experiment Evaluation Project (IRPhEP) provides students and young professionals the opportunity to gain experience and enhance critical engineering skills.

  4. Nuclear Technology Series. Course 24: Nuclear Systems and Safety.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  5. Nuclear Technology Series. Course 24: Nuclear Systems and Safety.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutions in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  6. Nuclear Technology Series. Course 8: Reactor Safety.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutians in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  7. Nuclear Technology Series. Course 8: Reactor Safety.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This technical specialty course is one of thirty-five courses designed for use by two-year postsecondary institutians in five nuclear technician curriculum areas: (1) radiation protection technician, (2) nuclear instrumentation and control technician, (3) nuclear materials processing technician, (4) nuclear quality-assurance/quality-control…

  8. Manned space flight nuclear system safety. Volume 1: Executive summary. Part 2: Space shuttle nuclear system safety

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The nuclear safety integration and operational aspects of transporting nuclear payloads to and from an earth orbiting space base by space shuttle are discussed. The representative payloads considered were: (1) zirconium hydride-Brayton power module, (2) isotope-Brayton power module, and (3) small isotope power systems or heat sources. Areas of investigation also include nuclear safety related integration and packaging as well as operational requirements for the shuttle and payload systems for all phases of the mission.

  9. Developing operational safety requirements for non-nuclear facilities

    SciTech Connect

    Mahn, J.A.

    1997-11-01

    Little guidance has been provided by the DOE for developing appropriate Operational Safety Requirements (OSR) for non-nuclear facility safety documents. For a period of time, Chapter 2 of DOE/AL Supplemental Order 5481.lB provided format guidance for non-reactor nuclear facility OSRs when this supplemental order applied to both nuclear and non-nuclear facilities. Thus, DOE Albuquerque Operations Office personnel still want to see non-nuclear facility OSRs in accordance with the supplemental order (i.e., in terms of Safety Limits, Limiting Conditions for Operation, and Administrative Controls). Furthermore, they want to see a clear correlation between the OSRs and the results of a facility safety analysis. This paper demonstrates how OSRs can be rather simply derived from the results of a risk assessment performed using the ``binning`` methodology of SAND95-0320.

  10. An Integrated Safety Assessment Methodology for Generation IV Nuclear Systems

    SciTech Connect

    Timothy J. Leahy

    2010-06-01

    The Generation IV International Forum (GIF) Risk and Safety Working Group (RSWG) was created to develop an effective approach for the safety of Generation IV advanced nuclear energy systems. Early work of the RSWG focused on defining a safety philosophy founded on lessons learned from current and prior generations of nuclear technologies, and on identifying technology characteristics that may help achieve Generation IV safety goals. More recent RSWG work has focused on the definition of an integrated safety assessment methodology for evaluating the safety of Generation IV systems. The methodology, tentatively called ISAM, is an integrated “toolkit” consisting of analytical techniques that are available and matched to appropriate stages of Generation IV system concept development. The integrated methodology is intended to yield safety-related insights that help actively drive the evolving design throughout the technology development cycle, potentially resulting in enhanced safety, reduced costs, and shortened development time.

  11. 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.

  12. Activities of the PNC Nuclear Safety Working Group

    SciTech Connect

    Kato, W.Y.

    1991-12-31

    The Nuclear Safety Working Group of the Pacific Nuclear Council promotes nuclear safety cooperation among its members. Status of safety research, emergency planning, development of lists of technical experts, severe accident prevention and mitigation have been the topics of discussion in the NSWG. This paper reviews and compares the severe accident prevention and mitigation program activities in some of the areas of the Pacific Basin region based on papers presented at a special session organized by the NSWG at an ANS Topical Meeting as well as papers from other sources.

  13. 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.

  14. Human Factors Research and Nuclear Safety.

    ERIC Educational Resources Information Center

    Moray, Neville P., Ed.; Huey, Beverly M., Ed.

    The Panel on Human Factors Research Needs in Nuclear Regulatory Research was formed by the National Research Council in response to a request from the Nuclear Regulatory Commission (NRC). The NRC asked the research council to conduct an 18-month study of human factors research needs for the safe operation of nuclear power plants. This report…

  15. 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.

  16. Nuclear safety, Volume 38, Number 1, January--March 1997

    SciTech Connect

    1997-03-01

    This journal contains nine articles which fall under the following categories: (1) general safety considerations; (2) control and instrumentation; (3) design features (4) environmental effects; (5) US Nuclear Regulatory Commission information and analyses; and (6) recent developments.

  17. A Safer Nuclear Enterprise - Application to Nuclear Explosive Safety (NES)(U)

    SciTech Connect

    Morris, Tommy J.

    2012-07-05

    Activities and infrastructure that support nuclear weapons are facing significant challenges. Despite an admirable record and firm commitment to make safety a primary criterion in weapons design, production, handling, and deployment - there is growing apprehension about terrorist acquiring weapons or nuclear material. At the NES Workshop in May 2012, Scott Sagan, who is a proponent of the normal accident cycle, presented. Whether a proponent of the normal accident cycle or High Reliability Organizations - we have to be diligent about our safety record. Constant vigilance is necessary to maintain our admirable safety record and commitment to Nuclear Explosive Safety.

  18. Nuclear energy safety challenges in the former Soviet Union

    SciTech Connect

    1995-12-31

    Fifteen nuclear reactors of the type that exploded at Chernobyl in April 1986 are still operating in Russia, Ukraine, and Lithuania. The West, concerned about safety of operations, wants these reactors shut down, but the host nations refuse. The electricity these reactors supply is nuch too important for their economies, so the argument goes. The report defines policy options and procedures to implement those options for the acceptable resolution of the nuclear power safety issues facing the former Soviet Union.

  19. Safety Oversight of Decommissioning Activities at DOE Nuclear Sites

    SciTech Connect

    Zull, Lawrence M.; Yeniscavich, William

    2008-01-15

    The Defense Nuclear Facilities Safety Board (Board) is an independent federal agency established by Congress in 1988 to provide nuclear safety oversight of activities at U.S. Department of Energy (DOE) defense nuclear facilities. The activities under the Board's jurisdiction include the design, construction, startup, operation, and decommissioning of defense nuclear facilities at DOE sites. This paper reviews the Board's safety oversight of decommissioning activities at DOE sites, identifies the safety problems observed, and discusses Board initiatives to improve the safety of decommissioning activities at DOE sites. The decommissioning of former defense nuclear facilities has reduced the risk of radioactive material contamination and exposure to the public and site workers. In general, efforts to perform decommissioning work at DOE defense nuclear sites have been successful, and contractors performing decommissioning work have a good safety record. Decommissioning activities have recently been completed at sites identified for closure, including the Rocky Flats Environmental Technology Site, the Fernald Closure Project, and the Miamisburg Closure Project (the Mound site). The Rocky Flats and Fernald sites, which produced plutonium parts and uranium materials for defense needs (respectively), have been turned into wildlife refuges. The Mound site, which performed R and D activities on nuclear materials, has been converted into an industrial and technology park called the Mound Advanced Technology Center. The DOE Office of Legacy Management is responsible for the long term stewardship of these former EM sites. The Board has reviewed many decommissioning activities, and noted that there are valuable lessons learned that can benefit both DOE and the contractor. As part of its ongoing safety oversight responsibilities, the Board and its staff will continue to review the safety of DOE and contractor decommissioning activities at DOE defense nuclear sites.

  20. THE IMPACT OF THE GLOBAL NUCLEAR SAFETY REGIME IN BRAZIL

    SciTech Connect

    Almeida, C.

    2004-10-06

    A turning point of the world nuclear industry with respect to safety occurred due to the accident at Chernobyl, in 1986. A side from the tragic personal losses and the enormous financial damage, the Chernobyl accident has literally demonstrated that ''a nuclear accident anywhere is an accident everywhere''. The impact was felt immediately by the nuclear industry, with plant cancellations (e.g. Austria), elimination of national programs (e.g. Italy) and general construction delays. However, the reaction of the nuclear industry was equally immediate, which led to the proposal and establishment of a Global Nuclear Safety Regime. This regime is composed of biding international safety conventions, globally accepted safety standard, and a voluntary peer review system. In a previous work, the author has presented in detail the components of this Regime, and briefly discussed its impact in the Brazilian nuclear power organizations, including the Regulatory Body. This work, on the opposite, briefly reviews the Global Nuclear Safety Regime, and concentrates in detail in the discussion of its impact in Brazil, showing how it has produced some changes, and where the peer pressure regime has failed to produce real results.

  1. Power plant of high safety for underground nuclear power station

    SciTech Connect

    Dolgov, V.N.

    1993-12-31

    An ecologically pure, reliable, and economic nuclear power station is based on the use of nuclear power plants with the liquid-metal coolant. This plant with the inherent safety is protected from external influences due to the underground accommodations in geologically stable formations such as granites, cambrian clays, and salt deposits. The design features of this underground plant are described.

  2. Government: Nuclear Safety in Doubt a Year after Accident.

    ERIC Educational Resources Information Center

    Ember, Lois R.

    1980-01-01

    A year after the accident at Three Mile Island (TMI), the signals transmitted to the public are still confused. Industry says that nuclear power is safe and that the aftermath of TMI ushers in a new era of safety. Antinuclear activists say TMI sounded nuclear power's death knell. (Author/RE)

  3. Government: Nuclear Safety in Doubt a Year after Accident.

    ERIC Educational Resources Information Center

    Ember, Lois R.

    1980-01-01

    A year after the accident at Three Mile Island (TMI), the signals transmitted to the public are still confused. Industry says that nuclear power is safe and that the aftermath of TMI ushers in a new era of safety. Antinuclear activists say TMI sounded nuclear power's death knell. (Author/RE)

  4. Martin Marietta Energy Systems Nuclear Criticality Safety Improvement Program

    SciTech Connect

    Speas, I.G.

    1987-04-29

    This report addresses questions raised by criticality safety violation at several DOE plants. Two charts are included that define the severity and reporting requirements for the six levels of accidents. A summary is given of all reported criticality incident at the DOE plants involved. The report concludes with Martin Marietta's Nuclear Criticality Safety Policy Statement. (JDH)

  5. The unique signal concept for detonation safety in nuclear weapons

    SciTech Connect

    Spray, S.D.; Cooper, J.A.

    1993-06-01

    The purpose of a unique signal (UQS) in a nuclear weapon system is to provide an unambiguous communication of intent to detonate from the UQS information input source device to a stronglink safety device in the weapon in a manner that is highly unlikely to be duplicated or simulated in normal environments and in a broad range of ill-defined abnormal environments. This report presents safety considerations for the design and implementation of UQSs in the context of the overall safety system.

  6. Current trends in nuclear safety programs at Brookhaven National Laboratory

    SciTech Connect

    Bari, R.A.; Duffey, R.B.; Baron, S.

    1993-12-31

    Brookhaven National Laboratory conducts nuclear safety research and technical assistance programs for the U.S. nuclear regulatory commission and for the Department of Energy. This includes experimental and analytical studies in the following areas: risk assessment associated with low power and shutdown of Pressurized water reactors (PWR`S); development of guidelines for accidental management related to containment and radiological releases; experiments on hydrogen combustion; plant aging and life extension; human reliability and management factors related to safety; reactor safety assessment of advanced reactor concepts; reactor physics analysis; structural analysis; and radiation protection of workers.

  7. Aerospace nuclear safety: An introduction and historical overview

    SciTech Connect

    Lee, J.H.; Buden, D.

    1994-04-01

    This paper provides an introduction and overview on the topical area of aerospace nuclear safety. Emphasis is on the history of the use of nuclear power sources in space, operational experience with these nuclear sources, a review of previous accidents associated with both U.S. and Russian launches, and the safety issues associated with the entire life cycle of space reactors. There are several potential missions to include near earth orbit, orbit-raising, lunar bases, and propulsion to such solar system locations as Mars, which are suitable for the use of space reactors. The process by which approval is obtained to launch these nuclear materials to space is also presented as well as the role of nuclear safety policy and requirements in a space program using nuclear power sources. Important differences in safety concerns for the Radioisotope Thermoelectric Generators (RTGs) now used, and space reactors are presented. The role and purpose of independent safety evaluation and assessment in ensuring safe launch and operation is also discussed. In summary, this paper provides the requisite framework in this topical area for the remaining papers of this session.

  8. NUCLEAR NONPROLIFERATION AND SAFETY: Challenges Facing the International Atomic Energy Agency.

    DTIC Science & Technology

    2007-11-02

    safeguards), and the Chernobyl nuclear power plant accident have focused greater attention on nuclear proliferation and the safety of nuclear power... Chernobyl , IAEA has placed increasing emphasis on assisting member states in improving the safety of nuclear power plants. Despite funding shortfalls...report language, GAO has incorporated their comments where appropriate. 2Nuclear Power Safety: Chernobyl Accident Prompted Worldwide Actions but

  9. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 3: Nuclear Safety Analysis Document (NSAD)

    NASA Technical Reports Server (NTRS)

    1972-01-01

    Nuclear safety analysis as applied to a space base mission is presented. The nuclear safety analysis document summarizes the mission and the credible accidents/events which may lead to nuclear hazards to the general public. The radiological effects and associated consequences of the hazards are discussed in detail. The probability of occurrence is combined with the potential number of individuals exposed to or above guideline values to provide a measure of accident and total mission risk. The overall mission risk has been determined to be low with the potential exposure to or above 25 rem limited to less than 4 individuals per every 1000 missions performed. No radiological risk to the general public occurs during the prelaunch phase at KSC. The most significant risks occur from prolonged exposure to reactor debris following land impact generally associated with the disposal phase of the mission where fission product inventories can be high.

  10. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS ENVIRONMENT, ENERGY AND WATER EFFICIENCY, RENEWABLE ENERGY TECHNOLOGIES, OCCUPATIONAL SAFETY, AND DRUG-FREE WORKPLACE Environmental, Energy and Water Efficiency, Renewable Energy Technologies, and Occupational...

  11. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS ENVIRONMENT, ENERGY AND WATER EFFICIENCY, RENEWABLE ENERGY TECHNOLOGIES, OCCUPATIONAL SAFETY, AND DRUG-FREE WORKPLACE Environmental, Energy and Water Efficiency, Renewable Energy Technologies, and Occupational...

  12. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS ENVIRONMENT, ENERGY AND WATER EFFICIENCY, RENEWABLE ENERGY TECHNOLOGIES, OCCUPATIONAL SAFETY, AND DRUG-FREE WORKPLACE Environmental, Energy and Water Efficiency, Renewable Energy Technologies, and Occupational...

  13. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS ENVIRONMENT, ENERGY AND WATER EFFICIENCY, RENEWABLE ENERGY TECHNOLOGIES, OCCUPATIONAL SAFETY, AND DRUG-FREE WORKPLACE Environmental, Energy and Water Efficiency, Renewable Energy Technologies, and Occupational...

  14. Nuclear Data Activities in Support of the DOE Nuclear Criticality Safety Program

    NASA Astrophysics Data System (ADS)

    Westfall, R. M.; McKnight, R. D.

    2005-05-01

    The DOE Nuclear Criticality Safety Program (NCSP) provides the technical infrastructure maintenance for those technologies applied in the evaluation and performance of safe fissionable-material operations in the DOE complex. These technologies include an Analytical Methods element for neutron transport as well as the development of sensitivity/uncertainty methods, the performance of Critical Experiments, evaluation and qualification of experiments as Benchmarks, and a comprehensive Nuclear Data program coordinated by the NCSP Nuclear Data Advisory Group (NDAG). The NDAG gathers and evaluates differential and integral nuclear data, identifies deficiencies, and recommends priorities on meeting DOE criticality safety needs to the NCSP Criticality Safety Support Group (CSSG). Then the NDAG identifies the required resources and unique capabilities for meeting these needs, not only for performing measurements but also for data evaluation with nuclear model codes as well as for data processing for criticality safety applications. The NDAG coordinates effort with the leadership of the National Nuclear Data Center, the Cross Section Evaluation Working Group (CSEWG), and the Working Party on International Evaluation Cooperation (WPEC) of the OECD/NEA Nuclear Science Committee. The overall objective is to expedite the issuance of new data and methods to the DOE criticality safety user. This paper describes these activities in detail, with examples based upon special studies being performed in support of criticality safety for a variety of DOE operations.

  15. Proceedings of the Nuclear Criticality Technology Safety Workshop

    SciTech Connect

    Rene G. Sanchez

    1998-04-01

    This document contains summaries of most of the papers presented at the 1995 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 16 and 17 at San Diego, Ca. The meeting was broken up into seven sessions, which covered the following topics: (1) Criticality Safety of Project Sapphire; (2) Relevant Experiments For Criticality Safety; (3) Interactions with the Former Soviet Union; (4) Misapplications and Limitations of Monte Carlo Methods Directed Toward Criticality Safety Analyses; (5) Monte Carlo Vulnerabilities of Execution and Interpretation; (6) Monte Carlo Vulnerabilities of Representation; and (7) Benchmark Comparisons.

  16. 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.

  17. Web-based nuclear criticality safety bibliographic database

    SciTech Connect

    Koponen, B L; Huang, S T

    2000-06-21

    The Lawrence Livermore National Laboratory has prepared a Nuclear Criticality Safety Bibliographic Database that is now available via the Internet. This database is a component of the U.S. DOE Nuclear Criticality Safety Program (NCSP) Web site. This WWW resource was developed as part of the DOE response to the DNFSB Recommendation 97-2, which reflected the need to make criticality safety information available to a wide audience. To the extent possible, the hyperlinks on the Web pages direct the user to original source of the reference material in order to ensure accuracy and access to the latest versions. A master index is in place for simple navigation through the site. A search capability is available to assist in locating the on-line reference materials. Among the features included are: A user-friendly site map for ease of use; A personnel registry; Links to all major laboratories and organizations involved in the many aspects of criticality safety; General help for new criticality safety practitioners, including basic technical references and training modules; A discussion of computational methods; An interactive question and answer forum for the criticality safety community; and Collections of bibliographic references mdvahdation experiments. This paper will focus on the bibliographic database. This database evolved from earlier work done by the DOE's Nuclear Criticality Information System (NCIS) maintained at LLNL during the 1980s. The bibliographic database at the time of the termination of NCIS were composed principally of three parts: (1) A critical experiment bibliography of 1067 citations (reported in UCRL-52769); (2) A compilation of criticality safety papers from Volumes 1 through 41 of the Transactions of the American Nuclear Society (reported in UCRL-53369); and (3) A general criticality bibliography of several thousand citations (unpublished). When the NCIS project was terminated the database was nearly lost but, fortunately, several years later

  18. Space Nuclear Safety Program. Progress report

    SciTech Connect

    Bronisz, S.E.

    1984-01-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

  19. Safety management of nuclear waste in Spain

    SciTech Connect

    Echavarri, L.E. )

    1991-01-01

    For the past two decades, Spain has been consolidating a nuclear program that in the last 3 years has provided between 35 and 40% of the electricity consumed in that country. This program includes nine operating reactor units, eight of them based on US technology and one from Germany, a total of 7,356 MW(electric). There is also a 480-MW(electric) French gas-cooled reactor whose operation recently ceased and which will be decommissioned in the coming years. Spanish industry has participated significantly in this program, and material produced locally has reached 85% of the total. Once the construction program has been completed and operation is proceeding normally, the capacity factor will be {approximately} 80%. It will be very important to complete the nuclear program with the establishment of conditions for safe management and disposal of the nuclear waste generated during the years in which these reactors are in operation and for subsequent decommissioning. To establish the guidelines for the disposal of nuclear waste, the Spanish government approved in october 1987, with a revision in January 1989, the General Plan of Radioactive Wastes proposed by the Ministry of Industry and Energy and prepared by the national company for radioactive waste management, ENRESA.

  20. Aging of safety class 1E transformers in safety systems of nuclear power plants

    SciTech Connect

    Roberts, E.W.; Edson, J.L.; Udy, A.C.

    1996-02-01

    This report discusses aging effects on safety-related power transformers in nuclear power plants. It also evaluates maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission`s (NRC`s) Nuclear Plant-Aging Research approach. It investigates the materials used in transformer construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes transformer failure events reported in various databases, and evaluates maintenance practices. Databases maintained by the nuclear industry were analyzed to evaluate the effects of aging on the operation of nuclear power plants.

  1. Guidance for identifying, reporting and tracking nuclear safety noncompliances

    SciTech Connect

    1995-12-01

    This document provides Department of Energy (DOE) contractors, subcontractors and suppliers with guidance in the effective use of DOE`s Price-Anderson nuclear safety Noncompliance Tracking System (NTS). Prompt contractor identification, reporting to DOE, and correction of nuclear safety noncompliances provides DOE with a basis to exercise enforcement discretion to mitigate civil penalties, and suspend the issuance of Notices of Violation for certain violations. Use of this reporting methodology is elective by contractors; however, this methodology is intended to reflect DOE`s philosophy on effective identification and reporting of nuclear safety noncompliances. To the extent that these expectations are met for particular noncompliances, DOE intends to appropriately exercise its enforcement discretion in considering whether, and to what extent, to undertake enforcement action.

  2. Nuclear safety. Volume 36, Number 2, July--December 1995

    SciTech Connect

    1995-12-01

    The primary scope of the journal is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. The following subjects are covered here: (1) the Chernobyl accident; (2) general safety considerations; (3) accident analysis; (4) design features; (5) environmental effects; (6) operating experiences; (7) US NRC information and analyses; and (8) recent developments. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  3. Nuclear Criticality Safety Organization qualification program. Revision 4

    SciTech Connect

    Carroll, K.J.; Taylor, R.G.; Worley, C.A.

    1997-05-19

    The Nuclear Criticality Safety Organization (NCSO) is committed to developing and maintaining a staff of highly qualified personnel to meet the current and anticipated needs in Nuclear Criticality Safety (NCS) at the Oak Ridge Y-12 Plant. This document defines the Qualification Program to address the NCSO technical and managerial qualification as required by the Y-12 Training Implementation Matrix (TIM). It is implemented through a combination of LMES plant-wide training courses and professional nuclear criticality safety training provided within the organization. This Qualification Program is applicable to technical and managerial NCSO personnel, including temporary personnel, sub-contractors and/or LMES employees on loan to the NCSO, who perform the NCS tasks or serve NCS-related positions as defined in sections 5 and 6 of this program.

  4. Nuclear space power safety and facility guidelines study

    SciTech Connect

    Mehlman, W.F.

    1995-09-11

    This report addresses safety guidelines for space nuclear reactor power missions and was prepared by The Johns Hopkins University Applied Physics Laboratory (JHU/APL) under a Department of Energy grant, DE-FG01-94NE32180 dated 27 September 1994. This grant was based on a proposal submitted by the JHU/APL in response to an {open_quotes}Invitation for Proposals Designed to Support Federal Agencies and Commercial Interests in Meeting Special Power and Propulsion Needs for Future Space Missions{close_quotes}. The United States has not launched a nuclear reactor since SNAP 10A in April 1965 although many Radioisotope Thermoelectric Generators (RTGs) have been launched. An RTG powered system is planned for launch as part of the Cassini mission to Saturn in 1997. Recently the Ballistic Missile Defense Office (BMDO) sponsored the Nuclear Electric Propulsion Space Test Program (NEPSTP) which was to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source in space. As of late 1993 the flight portion of this program was canceled but work to investigate the attributes of the reactor were continued but at a reduced level. While the future of space nuclear power systems is uncertain there are potential space missions which would require space nuclear power systems. The differences between space nuclear power systems and RTG devices are sufficient that safety and facility requirements warrant a review in the context of the unique features of a space nuclear reactor power system.

  5. Review of Overall Safety Manual for space nuclear systems. An evaluation of a nuclear safety analysis methodology for plutonium-fueled space nuclear systems

    SciTech Connect

    Coleman, J.; Inhaber, H.

    1984-02-01

    As part of its duties in connection with space missions involving nuclear power sources, the Office of Nuclear Safety (ONS) of the Office of Assistant Secretary for Environmental Protection, Safety, and Emergency Preparedness has been assigned the task of reviewing the Overall Safety Manual (OSM) (memo from B.J. Rock to J.R. Maher, December 1, 1982). The OSM, dated July 1981 and in four volumes, was prepared by NUS Corporation, Rockville, Maryland, for the US Department of Energy. The OSM provides many of the technical models and much of the data which are used by (1) space launch contractors in safety analysis reports and (2) the broader Interagency Nuclear Safety Review Panel (INSRP) safety evaluation reports. If fhs interaction between the OSM, contractors, and INSRP is to work effectively, the OSM must be accurate, comprehensive, understandable, and usable.

  6. Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities

    SciTech Connect

    Garvin, L.J.

    1995-11-01

    This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

  7. Perspectives of The Interagency Nuclear Safety Review Panel (INSRP) on future nuclear powered space missions

    SciTech Connect

    Gray, L.B. ); Pyatt, D.W. ); Sholtis, J.A. ); Winchester, R.O. , c/o Directorate of Nuclear Surety, Kirtland AFB, New Mexico 87117 )

    1993-01-10

    The Interagency Nuclear Safety Review Panel (INSRP) has provided reviews of all nuclear powered spacecraft launched by the United States. The two most recent launches were Ulysses in 1990 and Galileo in 1989. One reactor was launched in 1965 (SNAP-10A). All other U.S. space missions have utilized radioisotopic thermoelectric generators (RTGs). There are several missions in the next few years that are to be nuclear powered, including one that would utilize the Topaz II reactor purchased from Russia. INSRP must realign itself to perform parallel safety assessments of a reactor powered space mission, which has not been done in about thirty years, and RTG powered missions.

  8. 78 FR 4477 - Review of Safety Analysis Reports for Nuclear Power Plants, Introduction

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-22

    ... COMMISSION Review of Safety Analysis Reports for Nuclear Power Plants, Introduction AGENCY: Nuclear... subsection to NUREG-0800, ``Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power..., Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:...

  9. Training of nuclear criticality safety engineers

    SciTech Connect

    Taylor, R.G.

    1997-06-01

    The site specific analysis of nuclear criticality training needs is very briefly described. Analysis indicated that the four major components required were analysis, surveillance, business practices or administration, and emergency preparedness. The analysis component was further divided into process analysis, accident analysis, and transportation analysis. Ten subject matter areas for the process analysis component were identified as candidates for class development. Training classes developed from the job content analysis have demonstrated that the specialized information can be successfully delivered to new entrants. 1 fig.

  10. 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.

  11. Safety system augmentation at Russian nuclear power plants

    SciTech Connect

    Scerbo, J.A.; Satpute, S.N.; Donkin, J.Y.; Reister, R.A. |

    1996-12-31

    This paper describes the design and procurement of a Class IE DC power supply system to upgrade plant safety at the Kola Nuclear Power Plant (NPP). Kola NPP is located above the Arctic circle at Polyarnie Zorie, Murmansk, Russia. Kola NPP consists of four units. Units 1 and 2 have VVER-440/230 type reactors: Units 3 and 4 have VVER-440/213 type reactors. The VVER-440 reactor design is similar to the pressurized water reactor design used in the US. This project provided redundant, Class 1E DC station batteries and DC switchboards for Kola NPP, Units 1 and 2. The new DC power supply system was designed and procured in compliance with current nuclear design practices and requirements. Technical issues that needed to be addressed included reconciling the requirements in both US and Russian codes and satisfying the requirements of the Russian nuclear regulatory authority. Close interface with ATOMENERGOPROEKT (AEP), the Russian design organization, KOLA NPP plant personnel, and GOSATOMNADZOR (GAN), the Russian version of US Nuclear Regulatory Commission, was necessary to develop a design that would assure compliance with current Russian design requirements. Hence, this project was expected to serve as an example for plant upgrades at other similar VVER-440 nuclear plants. In addition to technical issues, the project needed to address language barriers and the logistics of shipping equipment to a remote section of the Former Soviet Union (FSU). This project was executed by Burns and Roe under the sponsorship of the US DOE as part of the International Safety Program (INSP). The INSP is a comprehensive effort, in cooperation with partners in other countries, to improve nuclear safety worldwide. A major element within the INSP is the improvement of the safety of Soviet-designed nuclear reactors.

  12. MOX LTA Fuel Cycle Analyses: Nuclear and Radiation Safety

    SciTech Connect

    Pavlovitchev, A.M.

    2001-09-28

    Tasks of nuclear safety assurance for storage and transport of fresh mixed uranium-plutonium fuel of the VVER-1000 reactor are considered in the view of 3 MOX LTAs introduction into the core. The precise code MCU that realizes the Monte Carlo method is used for calculations.

  13. Automating Nuclear-Safety-Related SQA Procedures with Custom Applications

    SciTech Connect

    Freels, James D.

    2016-01-01

    Nuclear safety-related procedures are rigorous for good reason. Small design mistakes can quickly turn into unwanted failures. Researchers at Oak Ridge National Laboratory worked with COMSOL to define a simulation app that automates the software quality assurance (SQA) verification process and provides results in less than 24 hours.

  14. Automating Nuclear-Safety-Related SQA Procedures with Custom Applications

    DOE PAGES

    Freels, James D.

    2016-01-01

    Nuclear safety-related procedures are rigorous for good reason. Small design mistakes can quickly turn into unwanted failures. Researchers at Oak Ridge National Laboratory worked with COMSOL to define a simulation app that automates the software quality assurance (SQA) verification process and provides results in less than 24 hours.

  15. Information Scanning and Processing at the Nuclear Safety Information Center.

    ERIC Educational Resources Information Center

    Parks, Celia; Julian, Carol

    This report is a detailed manual of the information specialist's duties at the Nuclear Safety Information Center. Information specialists scan the literature for documents to be reviewed, procure the documents (books, journal articles, reports, etc.), keep the document location records, and return the documents to the plant library or other…

  16. Safety aspects of nuclear waste disposal in space

    NASA Technical Reports Server (NTRS)

    Rice, E. E.; Edgecombe, D. S.; Compton, P. R.

    1981-01-01

    Safety issues involved in the disposal of nuclear wastes in space as a complement to mined geologic repositories are examined as part of an assessment of the feasibility of nuclear waste disposal in space. General safety guidelines for space disposal developed in the areas of radiation exposure and shielding, containment, accident environments, criticality, post-accident recovery, monitoring systems and isolation are presented for a nuclear waste disposal in space mission employing conventional space technology such as the Space Shuttle. The current reference concept under consideration by NASA and DOE is then examined in detail, with attention given to the waste source and mix, the waste form, waste processing and payload fabrication, shipping casks and ground transport vehicles, launch site operations and facilities, Shuttle-derived launch vehicle, orbit transfer vehicle, orbital operations and space destination, and the system safety aspects of the concept are discussed for each component. It is pointed out that future work remains in the development of an improved basis for the safety guidelines and the determination of the possible benefits and costs of the space disposal option for nuclear wastes.

  17. Space Nuclear Safety Program. Progress report, April 1984

    SciTech Connect

    George, T.G.

    1985-10-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Covered are: general-purpose heat source testing and recovery, and safety technology program (biaxial testing, iridium chemistry).

  18. Proceedings of the Nuclear Criticality Technology and Safety Project Workshop

    SciTech Connect

    Sanchez, R.G.

    1994-01-01

    This report is the proceedings of the annual Nuclear Criticality Technology and Safety Project (NCTSP) Workshop held in Monterey, California, on April 16--28, 1993. The NCTSP was sponsored by the Department of Energy and organized by the Los Alamos Critical Experiments Facility. The report is divided into six sections reflecting the sessions outlined on the workshop agenda.

  19. Safety aspects of nuclear waste disposal in space

    NASA Technical Reports Server (NTRS)

    Rice, E. E.; Edgecombe, D. S.; Compton, P. R.

    1981-01-01

    Safety issues involved in the disposal of nuclear wastes in space as a complement to mined geologic repositories are examined as part of an assessment of the feasibility of nuclear waste disposal in space. General safety guidelines for space disposal developed in the areas of radiation exposure and shielding, containment, accident environments, criticality, post-accident recovery, monitoring systems and isolation are presented for a nuclear waste disposal in space mission employing conventional space technology such as the Space Shuttle. The current reference concept under consideration by NASA and DOE is then examined in detail, with attention given to the waste source and mix, the waste form, waste processing and payload fabrication, shipping casks and ground transport vehicles, launch site operations and facilities, Shuttle-derived launch vehicle, orbit transfer vehicle, orbital operations and space destination, and the system safety aspects of the concept are discussed for each component. It is pointed out that future work remains in the development of an improved basis for the safety guidelines and the determination of the possible benefits and costs of the space disposal option for nuclear wastes.

  20. Integrated deterministic and probabilistic safety analysis for safety assessment of nuclear power plants

    DOE PAGES

    Di Maio, Francesco; Zio, Enrico; Smith, Curtis; ...

    2015-07-06

    The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs andmore » activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).« less

  1. Integrated deterministic and probabilistic safety analysis for safety assessment of nuclear power plants

    SciTech Connect

    Di Maio, Francesco; Zio, Enrico; Smith, Curtis; Rychkov, Valentin

    2015-07-06

    The present special issue contains an overview of the research in the field of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA) of Nuclear Power Plants (NPPs). Traditionally, safety regulation for NPPs design and operation has been based on Deterministic Safety Assessment (DSA) methods to verify criteria that assure plant safety in a number of postulated Design Basis Accident (DBA) scenarios. Referring to such criteria, it is also possible to identify those plant Structures, Systems, and Components (SSCs) and activities that are most important for safety within those postulated scenarios. Then, the design, operation, and maintenance of these “safety-related” SSCs and activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).

  2. Passive Safety Features in Advanced Nuclear Power Plant Design

    NASA Astrophysics Data System (ADS)

    Tahir, M.; Chughtai, I. R.; Aslam, M.

    2013-03-01

    For implementation of advance passive safety features in future nuclear power plant design, a passive safety system has been proposed and its response has been observed for Loss of Coolant Accident (LOCA) in the cold leg of a reactor coolant system. In a transient simulation the performance of proposed system is validated against existing safety injection system for a reference power plant of 325 MWe. The existing safety injection system is a huge system and consists of many active components including pumps, valves, piping and Instrumentation and Control (I&C). A good running of the active components of this system is necessary for its functionality as high head safety injection system under design basis accidents. Using reactor simulation technique, the proposed passive safety injection system and existing safety injection system are simulated and tested for their performance under large break LOCA for the same boundary conditions. Critical thermal hydraulic parameters of both the systems are presented graphically and discussed. The results obtained are approximately the same in both the cases. However, the proposed passive safety injection system is a better choice for such type of reactors due to reduction in components with improved safety.

  3. Safety analysis of irradiated nuclear fuel transportation container

    SciTech Connect

    Uspuras, E.; Rimkevicius, S.

    2007-07-01

    Ignalina NPP comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit were available for further reuse in Unit 2. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 fuel for reuse in the Unit 2. The Safety Analysis Report (SAR) was developed to demonstrate that the proposed set of equipment performs all functions and assures the required level of safety for both normal operation and accident conditions. The purpose of this paper is to introduce the content and main results of SAR, focusing attention on the container used to transport spent fuel assemblies from Unit I on Unit 2. In the SAR, the structural integrity, thermal, radiological and nuclear safety calculations are performed to assess the acceptance of the proposed set of equipment. The safety analysis demonstrated that the proposed nuclear fuel transportation container and other equipment are in compliance with functional, design and regulatory requirements and assure the required safety level. (authors)

  4. 77 FR 1748 - Atomic Safety and Licensing Board; Calvert Cliffs 3 Nuclear Project, LLC, and UniStar Nuclear...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-11

    ... COMMISSION Atomic Safety and Licensing Board; Calvert Cliffs 3 Nuclear Project, LLC, and UniStar Nuclear... Calvert Cliffs 3 Nuclear Project, L.L.C., and UniStar Nuclear Operating Services, L.L.C. (Applicants) for... Citizens Alliance for Renewable Energy Solutions; (2) UniStar Nuclear Operating Services, LLC and...

  5. An Empirical Analysis of Human Performance and Nuclear Safety Culture

    SciTech Connect

    Jeffrey Joe; Larry G. Blackwood

    2006-06-01

    The purpose of this analysis, which was conducted for the US Nuclear Regulatory Commission (NRC), was to test whether an empirical connection exists between human performance and nuclear power plant safety culture. This was accomplished through analyzing the relationship between a measure of human performance and a plant’s Safety Conscious Work Environment (SCWE). SCWE is an important component of safety culture the NRC has developed, but it is not synonymous with it. SCWE is an environment in which employees are encouraged to raise safety concerns both to their own management and to the NRC without fear of harassment, intimidation, retaliation, or discrimination. Because the relationship between human performance and allegations is intuitively reciprocal and both relationship directions need exploration, two series of analyses were performed. First, human performance data could be indicative of safety culture, so regression analyses were performed using human performance data to predict SCWE. It also is likely that safety culture contributes to human performance issues at a plant, so a second set of regressions were performed using allegations to predict HFIS results.

  6. Validation of the group nuclear safety climate questionnaire.

    PubMed

    Navarro, M Felisa Latorre; Gracia Lerín, Francisco J; Tomás, Inés; Peiró Silla, José María

    2013-09-01

    Group safety climate is a leading indicator of safety performance in high reliability organizations. Zohar and Luria (2005) developed a Group Safety Climate scale (ZGSC) and found it to have a single factor. The ZGSC scale was used as a basis in this study with the researchers rewording almost half of the items on this scale, changing the referents from the leader to the group, and trying to validate a two-factor scale. The sample was composed of 566 employees in 50 groups from a Spanish nuclear power plant. Item analysis, reliability, correlations, aggregation indexes and CFA were performed. Results revealed that the construct was shared by each unit, and our reworded Group Safety Climate (GSC) scale showed a one-factor structure and correlated to organizational safety climate, formalized procedures, safety behavior, and time pressure. This validation of the one-factor structure of the Zohar and Luria (2005) scale could strengthen and spread this scale and measure group safety climate more effectively. Copyright © 2013 National Safety Council and Elsevier Ltd. All rights reserved.

  7. Natural Disasters and Safety Risks at Nuclear Power Stations

    NASA Astrophysics Data System (ADS)

    Tutnova, T.

    2012-04-01

    In the aftermath of Fukushima natural-technological disaster the global opinion on nuclear energy divided even deeper. While Germany, Italy and the USA are currently reevaluating their previous plans on nuclear growth, many states are committed to expand nuclear energy output. In China and France, where the industry is widely supported by policymakers, there is little talk about abandoning further development of nuclear energy. Moreover, China displays the most remarkable pace of nuclear development in the world: it is responsible for 40% of worldwide reactors under construction, and aims at least to quadruple its nuclear capacity by 2020. In these states the consequences of Fukushima natural-technological accident will probably result in safety checks and advancement of new reactor technologies. Thus, China is buying newer reactor design from the USA which relies on "passive safety systems". It means that emergency power generators, crucial for reactor cooling in case of an accident, won't depend on electricity, so that tsunami won't disable them like it happened in the case of Fukushima. Nuclear energy managed to draw lessons from previous nuclear accidents where technological and human factors played crucial role. But the Fukushima lesson shows that the natural hazards, nevertheless, were undervalued. Though the ongoing technological advancements make it possible to increase the safety of nuclear power plants with consideration of natural risks, it is not just a question of technology improvement. A necessary action that must be taken is the reevaluation of the character and sources of the potential hazards which natural disasters can bring to nuclear industry. One of the examples is a devastating impact of more than one natural disaster happening at the same time. This subject, in fact, was not taken into account before, while it must be a significant point in planning sites for new nuclear power plants. Another important lesson unveiled is that world nuclear

  8. Proceedings of the nuclear criticality technology safety project

    SciTech Connect

    Sanchez, R.G.

    1997-06-01

    This document contains summaries of the most of the papers presented at the 1994 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 10 and 11 at Williamsburg, Va. The meeting was broken up into seven sessions, which covered the following topics: (1) Validation and Application of Calculations; (2) Relevant Experiments for Criticality Safety; (3) Experimental Facilities and Capabilities; (4) Rad-Waste and Weapons Disassembly; (5) Criticality Safety Software and Development; (6) Criticality Safety Studies at Universities; and (7) Training. The minutes and list of participants of the Critical Experiment Needs Identification Workgroup meeting, which was held on May 9 at the same venue, has been included as an appendix. A second appendix contains the names and addresses of all NCTSP meeting participants. Separate abstracts have been indexed to the database for contributions to this proceedings.

  9. 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.

  10. HFE safety reviews of advanced nuclear power plant control rooms

    NASA Technical Reports Server (NTRS)

    Ohara, John

    1994-01-01

    Advanced control rooms (ACR's) will utilize human-system interface (HSI) technologies that may have significant implications for plant safety in that they will affect the operator's overall role and means of interacting with the system. The Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) aspects of HSI's to ensure that they are designed to good HFE principles and support performance and reliability in order to protect public health and safety. However, the only available NRC guidance was developed more than ten years ago, and does not adequately address the human performance issues and technology changes associated with ACR's. Accordingly, a new approach to ACR safety reviews was developed based upon the concept of 'convergent validity'. This approach to ACR safety reviews is described.

  11. History of US nuclear weapon safety assessment: The early years

    SciTech Connect

    Spray, S.D.

    1996-06-01

    From the beginnings of the U.S. nuclear weapons program, military and civilian dual- agency judgment has been fundamental to achieving nuclear weapon and weapon system safety. This interaction was initiated by the Atomic Energy Act of 1946, which created the Atomic Energy Commission (AEC). The principle of using dual-agency judgment has been perpetuated in the design and assessment of the weapon and weapon system acceptance process since that time. This fundamental approach is still used today in all phases of the weapon life. In this paper, an overview of the history and philosophy of the approach is described.

  12. Qualification of Safety-Related Software in Nuclear Power Plants

    SciTech Connect

    Johnson, G L

    2000-06-13

    Digital instrumentation and control systems have the potential of offering significant benefits over traditional analog systems in Nuclear Power Plant safety systems, but there are also significant difficulties in qualifying digital systems to the satisfaction of regulators. Digital systems differ in fundamental ways from analog systems. New methods for safety qualification, which take these differences into account, would ease the regulatory cost and promote use of digital systems. This paper offers a possible method for assisting in the analysis of digital system software, as one step in an improved qualification process.

  13. Safety assessment of a robotic system handling nuclear material

    SciTech Connect

    Atcitty, C.B.; Robinson, D.G.

    1996-02-01

    This paper outlines the use of a Failure Modes and Effects Analysis for the safety assessment of a robotic system being developed at Sandia National Laboratories. The robotic system, The Weigh and Leak Check System, is to replace a manual process at the Department of Energy facility at Pantex by which nuclear material is inspected for weight and leakage. Failure Modes and Effects Analyses were completed for the robotics process to ensure that safety goals for the system had been meet. These analyses showed that the risks to people and the internal and external environment were acceptable.

  14. Reevaluating nuclear safety and security in a post 9/11 era.

    SciTech Connect

    Booker, Paul M.; Brown, Lisa M.

    2005-07-01

    This report has the following topics: (1) Changing perspectives on nuclear safety and security; (2) Evolving needs in a post-9/11 era; (3) Nuclear Weapons--An attractive terrorist target; (4) The case for increased safety; (5) Evolution of current nuclear weapons safety and security; (6) Integrated surety; (7) The role of safety and security in enabling responsiveness; (8) Advances in surety technologies; and (9) Reevaluating safety.

  15. 77 FR 50727 - Configuration Management Plans for Digital Computer Software Used in Safety Systems of Nuclear...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-22

    ... in Safety Systems of Nuclear Power Plants.'' The DG-1206 is proposed Revision 1 of RG 1.169, dated... Digital Computer Software Used in Safety Systems of Nuclear Power Plants'' is temporarily identified by... Plans for Digital Computer Software Used in Safety Systems of Nuclear Power Plants'' dated September...

  16. Developing glovebox robotics to meet the national robot safety standard and nuclear safety criteria

    SciTech Connect

    McMahon, T.T.; Sievers, R.H. )

    1991-09-01

    Development of a glove box based robotic system by the Lawrence Livermore National Laboratory (LLNL) is reported. Safety issues addressed include planning to meet the special constraints of operations within a hazardous material glove box and with hostile environments, compliance with the current and draft national robotic system safety standards, and eventual satisfaction of nuclear material handling requirements. Special attention has been required for the revision to the robot and control system models which antedate adoption of the present national safety standard. A robotic test bed, using non-radioactive surrogates is being activated at the Lawrence Livermore National Laboratory to develop the material handling system and the process interfaces for future special nuclear material processing applications. Part of this effort is to define, test, and revise adequate safety controls to ensure success when the system is eventually deployed at a DOE site. The current system is primarily for demonstration and testing, but will evolve into the baseline configuration from which the production system is to be derived. This results in special hazards associated with research activities which may not be present on a production line. Nuclear safety is of paramount importance and has been successfully addressed for 50 years in the DOE weapons production complex. It carries its particular requirements for robot systems and manual operations, as summarized below: Criticality must be avoided (materials cannot consolidate or accumulate to approach a critical mass). Radioactive materials must be confined. The public and workers must be protected from accountable radiation exposure. Nuclear material must be readily retrievable. Nuclear safety must be conclusively demonstrated through hazards analysis. 7 refs.

  17. Nuclear Safety of RBMK Storage Pool under Seismic Impact

    NASA Astrophysics Data System (ADS)

    Fedosov, A.

    2017-01-01

    Nuclear safety of RBMK storage pool of spent fuel during of the maximum design earthquake is evaluated. The lower ends of the fuel assemblies are not fixed and they can deviate from the vertical position. The seismic action may be one of the reasons for such deviations. 3D model of fuel assemblies movements caused by seismic impact is used. The simulation of the dynamics of a fuel assemblies group under seismic impacts allows to find the dangerous configuration of closest approach of the fuel assemblies. Three-dimensional neutron program STEPAN calculates the Keff of the most dangerous systems. The maximum design earthquake is the design basis accident. In this case according to the regulatory documents the fuel is considered with zero burn-up. Nuclear safety of RBMK storage pool under considered conditions is provided.

  18. What You Should Know About Pediatric Nuclear Medicine and Radiation Safety

    MedlinePlus

    What You Should Know About Pediatric Nuclear Medicine and Radiation Safety www.imagegently.org What is nuclear medicine? Nuclear medicine uses radioactive isotopes to create pictures of the human body. These pictures ...

  19. 75 FR 50009 - Babcock & Wilcox Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-16

    ... COMMISSION Babcock & Wilcox Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing... & Wilcox Nuclear Operations Group, Inc. (Lynchburg, VA Facility). This proceeding concerns an Order Imposing Civil Monetary Penalty served upon the Licensee, Babcock & Wilcox Nuclear Operations Group,...

  20. Safety provisions for the SP-100 Nuclear Assembly Test article

    NASA Astrophysics Data System (ADS)

    Temme, Mark I.; Damon, Dennis R.; Hackford, Norman E.; Ha, Chuong T.; Mapes, Robert L.; Miller, David D.; Smith, Michael A.

    The SP-100 Nuclear Assembly Test (NAT) article is charged with the validation of such a fast-spectrum, liquid metal-cooled reactor's integrated reactor, control, and shield performance characteristics in a spacelike environment. The NAT facility is designed for safe assembly, operation, disassembly, and disposal. Attention is presently given to the selection and classification of postulated accidents, safety design criteria, reactor trip parameters, and the structure of accident analyses.

  1. Double-clad nuclear-fuel safety rod

    DOEpatents

    McCarthy, W.H.; Atcheson, D.B.

    1981-12-30

    A device for shutting down a nuclear reactor during an undercooling or overpower event, whether or not the reactor's scram system operates properly. This is accomplished by double-clad fuel safety rods positioned at various locations throughout the reactor core, wherein melting of a secondary internal cladding of the rod allows the fuel column therein to shift from the reactor core to place the reactor in a subcritical condition.

  2. Space Nuclear Safety Program. Progress report, November 1983

    SciTech Connect

    Bronisz, S.E.

    1984-06-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Topics discussed include: safety-verification impact tests; explosion test; fragment test; leaking fueled clads; effects of fresh water and seawater or PuO/sub 2/ pellets; and impact tests of 5 watt radioisotope thermoelectric generator.

  3. Merger of Nuclear Data with Criticality Safety Calculations

    SciTech Connect

    Derrien, H.; Larson, N.M.; Leal, L.C.

    1999-09-20

    In this paper we report on current activities related to the merger of differential/integral data (especially in the resolved-resonance region) with nuclear criticality safety computations. Techniques are outlined for closer coupling of many processes � measurement, data reduction, differential-data analysis, integral-data analysis, generating multigroup cross sections, data-testing, criticality computations � which in the past have been treated independently.

  4. A comparison of commercial/industry and nuclear weapons safety concepts

    SciTech Connect

    Bennett, R.R.; Summers, D.A.

    1996-07-01

    In this paper the authors identify factors which influence the safety philosophy used in the US commercial/industrial sector and compare them against those factors which influence nuclear weapons safety. Commercial/industrial safety is guided by private and public safety standards. Generally, private safety standards tend to emphasize product reliability issues while public (i.e., government) safety standards tend to emphasize human factors issues. Safety in the nuclear weapons arena is driven by federal requirements and memoranda of understanding (MOUs) between the Departments of Defense and Energy. Safety is achieved through passive design features integrated into the nuclear weapon. Though the common strand between commercial/industrial and nuclear weapons safety is the minimization of risk posed to the general population (i.e., public safety), the authors found that each sector tends to employ a different safety approach to view and resolve high-consequence safety issues.

  5. US support for nuclear energy safety and cooperation in the Pacific Basin

    SciTech Connect

    Selin, I.

    1994-12-31

    Nuclear power plays an important role in the energy and economic development of Pacific Rim countries. Concurrent with the construction of commercial nuclear power plants, there is a vital need to develop strong nuclear safety infrastructures in all countries choosing to use nuclear energy for electricity production. One of the most important elements in developing a viable nuclear program is a nuclear safety culture, rigorously applied to nuclear plant siting, design, construction, operation and management. International cooperation provides an important mechanism for raising the level of nuclear safety worldwide. The NRC has recently increased its international efforts in the nuclear safety area, with particular emphasis on Central and Eastern Europe and the New Independent States of the former Soviet Union. The NRC will continue, and is prepared to expand, its cooperation with counterpart organizations in Pacific Rim countries to help establish and nurture a safety culture that will respond to the dynamic process of nuclear energy development over the next few years.

  6. Work practices, fatigue, and nuclear power plant safety performance.

    PubMed

    Baker, K; Olson, J; Morisseau, D

    1994-06-01

    This paper focuses on work practices that may contribute to fatigue-induced performance decrements in the commercial nuclear power industry. Specifically, the amount of overtime worked by operations, technical, and maintenance personnel and the 12-h operator shift schedule are studied. Although overtime for all three job categories was fairly high at a number of plants, the analyses detected a clear statistical relationship only between operations overtime and plant safety performance. The results for the 12-h operator shift schedule were ambiguous. Although the 12-h operator shift schedule was correlated with operator error, it was not significantly related to the other five safety indicators. This research suggests that at least one of the existing work practices--the amount of operator overtime worked at some plants--represents a safety concern in this industry; however, further research is required before any definitive conclusions can be drawn.

  7. PRELIMINARY NUCLEAR CRITICALITY NUCLEAR SAFETY EVLAUATION FOR THE CONTAINER SURVEILLANCE AND STORAGE CAPABILITY PROJECT

    SciTech Connect

    Low, M; Matthew02 Miller, M; Thomas Reilly, T

    2007-04-30

    Washington Safety Management Solutions (WSMS) provides criticality safety services to Washington Savannah River Company (WSRC) at the Savannah River Site. One activity at SRS is the Container Surveillance and Storage Capability (CSSC) Project, which will perform surveillances on 3013 containers (hereafter referred to as 3013s) to verify that they meet the Department of Energy (DOE) Standard (STD) 3013 for plutonium storage. The project will handle quantities of material that are greater than ANS/ANSI-8.1 single parameter mass limits, and thus required a Nuclear Criticality Safety Evaluation (NCSE). The WSMS methodology for conducting an NCSE is outlined in the WSMS methods manual. The WSMS methods manual currently follows the requirements of DOE-O-420.1B, DOE-STD-3007-2007, and the Washington Savannah River Company (WSRC) SCD-3 manual. DOE-STD-3007-2007 describes how a NCSE should be performed, while DOE-O-420.1B outlines the requirements for a Criticality Safety Program (CSP). The WSRC SCD-3 manual implements DOE requirements and ANS standards. NCSEs do not address the Nuclear Criticality Safety (NCS) of non-reactor nuclear facilities that may be affected by overt or covert activities of sabotage, espionage, terrorism or other security malevolence. Events which are beyond the Design Basis Accidents (DBAs) are outside the scope of a double contingency analysis.

  8. Main contributions of the KfK nuclear safety project in the LWR safety area

    SciTech Connect

    Kuczera, B.

    1986-01-01

    The Nuclear Safety Project (PNS) was established at the Kernforschungszentrum Karlsruhe (KfK) in 1972. At that time, nuclear energy in the Federal Republic of Germany was in a transition phase proceeding from light water reactor (LWR) demonstration plants (300 MW(e)) to commercial size plants of 1200 to 1300 MW(e) which are standard units today. Simultaneously, general questions about LWR safety and reliability as well as questions on risk-oriented features became more pronounced in the public discussion. As a consequence, various already existing LWR safety activities were brought together and combined in the organizational framework of the PNS. The overriding objectives of PNS research and development (R and D) effort were at the quantification of safety margins of reactor systems and components, and the improvement of existing safety systems to avoid accident occurrence and to minimize accident consequences. In close cooperation with governmental authorities, manufacturers, and utilities, an R and D program was developed, comprised of four main areas: 1) dynamic behavior of reactor components; 2) fuel element behavior under accident conditions; 3) core meltdown accident analyses; and 4) retention of radioactive fission products and limitation of severe accident consequences. An overview on the KfK contribution to LWR safety research is given. It deals in a comprehensive matter with results obtained in the areas listed above.

  9. Safety system augmentation at Russian Nuclear Power Plants

    NASA Astrophysics Data System (ADS)

    Scerbo, J. A.; Satpute, S. N.; Donkin, J. Y.; Reister, R. A.

    1997-06-01

    This paper describes the design and procurement of a Class 1E DC power supply system to upgrade plant safety at the Kola Nuclear Power Plant (NPP). Kola NPP is located above the Arctic circle at Polyanie Zprie, Murmansk, Russia. Kola NPP consists of four units. Units 1 and 2 have VVER-440/230 type reactors: Units 3 and 4 have VVER-440/213 type reactors. The VVER 440 reactor design is similar to the pressurized water reactor design used in the US. This project provided redundant Class 1E DC station batteries and DC switchboards for Kola NPP, Units 1 and 2. The new DC power supply system was designed and procured in compliance with current nuclear design practices and requirements. Technical issues that needed to be addressed included reconciling the requirements in both US and Russian codes and satisfying the requirements of the Russian nuclear regulatory authority. Close interface with ATOMENERGOPROEKT (AEP), the Russian design organization, KOLA NPP plant personnel, and GOSATOMNASZOR (GAN), the Russian Version of US Nuclear Regulatory Commission was necessary to develop a design that would assure compliance with current Russian design requirements. Hence, this project was expected to serve as an example for plant upgrades at other similar VVER-440 nuclear plants. In addition to technical issues, the project needed to address language barriers and the logistics of shipping equipment to a remote section of the Former Soviet Union.

  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. Nuclear Safety. Technical Progress Journal, October--December 1992: Volume 33, No. 4

    SciTech Connect

    1992-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  12. Nuclear Safety. Technical Progress Journal, January--March 1993: Volume 34, No. 1

    SciTech Connect

    1993-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  13. Nuclear Safety. Technical progress journal, April--June 1996: Volume 37, No. 2

    SciTech Connect

    Muhlheim, M D

    1996-01-01

    This journal covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

  14. Nuclear Safety. Technical progress journal, April--June 1992: Volume 33, No.2

    SciTech Connect

    Silver, E G

    1992-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  15. Nuclear Safety. Technical Progress Journal, January--March 1992: Volume 33, No. 1

    SciTech Connect

    1992-01-01

    This review journal covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  16. Nuclear Safety. Technical progress journal, January--March 1994: Volume 35, No. 1

    SciTech Connect

    Silver, E G

    1994-01-01

    This is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities.

  17. NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

    NASA Technical Reports Server (NTRS)

    Marshall, Albert C.; Sawyer, J. C., Jr.; Bari, Robert A.; Brown, Neil W.; Cullingford, Hatice S.; Hardy, Alva C.; Lee, James H.; Mcculloch, William H.; Niederauer, George F.; Remp, Kerry

    1992-01-01

    An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top-level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of safety functional requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

  18. Integrated safety program for the Nuclear Electric Propulsion Space Test Program

    NASA Astrophysics Data System (ADS)

    Marshall, A. C.; Mehlman, W. F.; Kompanietz, G.

    The Nuclear Electric Propulsion Space Test Program (NEPSTP) is sponsored by the Ballistic Missile Defense Office (BMDO) to demonstrate and evaluate the Russian-built TOPAZ 2 nuclear reactor as a power source for an electric propulsion system in space. From its inception, safety has been a central feature of the NEPSTP program. This paper addresses the work done to define the safety organizational relationships, responsibilities, management, engineering requirements, and documentation to assure an integrated safety program that coordinates the various safety activities in Mission Safety, Range Safety and Nuclear Safety. Because the United States has not launched a nuclear reactor since 1965, much of the focus of the safety program has been directed toward the unique safety considerations of using a nuclear reactor in space. Our preliminary findings indicate that the safe use of the TOPAZ 2 for the NEPSTP space mission is feasible.

  19. Integrated Safety Program for the Nuclear Electric Propulsion Space Test Program

    NASA Astrophysics Data System (ADS)

    Marshall, Albert C.; Mehlman, William F.; Kompanietz, G.

    1994-07-01

    The Nuclear Electric Propulsion Space Test Program (NEPSTP) is sponsored by the Ballistic Missile Defense Office (BMDO) to demonstrate and evaluate the Russian-built TOPAZ II nuclear reactor as a power source for an electric propulsion system in space. From its inception, safety has been a central feature of the NEPSTP program. This paper addresses the work done to define the safety organizational relationships, responsibilities, management, engineering requirements, and documentation to assure an integrated safety program that coordinates the various safety activities in Mission Safety, Range Safety and Nuclear Safety. Because the United States has not launched a nuclear reactor since 1965, much of the focus of the safety program has been directed toward the unique safety considerations of using a nuclear reactor in space. Our preliminary findings indicate that the safe use of the TOPAZ II for the NEPSTP space mission is feasible.

  20. NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

    NASA Technical Reports Server (NTRS)

    Marshall, Albert C.; Sawyer, J. C., Jr.; Bari, Robert A.; Brown, Neil W.; Cullingford, Hatice S.; Hardy, Alva C.; Lee, James H.; Mcculloch, William H.; Niederauer, George F.; Remp, Kerry

    1992-01-01

    An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top-level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of safety functional requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

  1. An interagency space nuclear propulsion safety policy for SEI - Issues and discussion

    NASA Technical Reports Server (NTRS)

    Marshall, A. C.; Sawyer, J. C., Jr.

    1991-01-01

    An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top level safety requirements and guidelines to address these issues. Safety topics include reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations. In this paper the emphasis is placed on the safety policy and the issues and considerations that are addressed by the NSPWG recommendations.

  2. NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

    SciTech Connect

    Marshall, A.C.; Lee, J.H.; McCulloch, W.H. ); Sawyer, J.C. Jr. ); Bari, R.A. ); Brown, N.W. ); Cullingford, H.S.; Hardy, A.C. (National Aeronautics and Space Administ

    1992-01-01

    An Interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top- level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

  3. NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

    SciTech Connect

    Marshall, A.C.; Lee, J.H.; McCulloch, W.H.; Sawyer, J.C. Jr.; Bari, R.A.; Brown, N.W.; Cullingford, H.S.; Hardy, A.C.; Niederauer, G.F.; Remp, K.; Rice, J.W.; Sholtis, J.A.

    1992-09-01

    An Interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top- level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

  4. NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

    NASA Astrophysics Data System (ADS)

    Marshall, Albert C.; Sawyer, J. C., Jr.; Bari, Robert A.; Brown, Neil W.; Cullingford, Hatice S.; Hardy, Alva C.; Lee, James H.; McCulloch, William H.; Niederauer, George F.; Remp, Kerry

    1992-07-01

    An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top-level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of safety functional requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed.

  5. An interagency space nuclear propulsion safety policy for SEI - Issues and discussion

    NASA Technical Reports Server (NTRS)

    Marshall, A. C.; Sawyer, J. C., Jr.

    1991-01-01

    An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of Safety Functional Requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top level safety requirements and guidelines to address these issues. Safety topics include reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, safeguards, risk/reliability, operational safety, ground testing, and other considerations. In this paper the emphasis is placed on the safety policy and the issues and considerations that are addressed by the NSPWG recommendations.

  6. 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.

  7. (Safety and reliability of nuclear power plant technology)

    SciTech Connect

    Dickson, T.L.

    1990-10-22

    The traveler attended the 16th MPA Seminar on the Safety and Reliability of Plant Technology with Special Emphasis on Nuclear Technology. The objective of the trip was to gather information and data that could prove useful to the US Nuclear Regulatory Commission (USNRC) sponsored Heavy-Section Steel Irradiation (HSSI) and Heavy-Section Steel Technology (HSST) Programs and to present a paper entitled, Effects of Irradiation on Initiation and Crack-Arrest Toughness of Two High-Copper Welds and on Stainless Steel Cladding. This paper summarizes results from the 5th, 6th, and 7th Irradiation Series of experiments performed within the HSSI Program by the Metals and Ceramics Division at Oak Ridge National Laboratory (ORNL).

  8. Nuclear reactor safety research since Three Mile Island

    SciTech Connect

    Mynatt, F.R.

    1982-04-09

    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.

  9. Nuclear reactor safety research since three mile island.

    PubMed

    Mynatt, F R

    1982-04-09

    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.

  10. Nuclear criticality safety for drums at Babcock and Wilcox

    SciTech Connect

    Alcorn, F.M.

    1997-12-01

    The Babcock and Wilcox Company (B&W) operates a nuclear fuel facility in Lynchburg, Virginia, processing uranium over the full range of possible enrichments (depleted to 97.65 wt% {sup 235}U). Nuclear fuel is produced for defense programs and for various research and test reactors worldwide. The facility has a uranium recovery operation that handles scrap produced at B&W as well as scrap from other U.S. Department of Energy sites. B&W also has a down-blending operation that is currently completing the down-blending of the fully enriched Project Sapphire Uranium to commercial-grade fuel (4 Wt% {sup 235}U). The facility generates approximately two hundred 55-gal drums of radioactive waste each month. Just a few years ago the number of waste drums on-site stood at {approximately}5000; however, through an aggressive waste reduction program, this number has been reduced to {approximately}2000. B&W strives to avoid storing uranium scrap in 55-gal drums; however, there are approximately sixty-four 55-gal drums of scrap on-site. Scrap is that material from which the uranium is recovered because of financial, contractual, or regulatory considerations; waste is that material destined for disposal. Whether waste or scrap, nuclear criticality safety is of paramount concern in the handling, processing, and storing of uranium-bearing drums at B&W. Any shipment complies with applicable U.S. Nuclear Regulatory Commission and U.S. Department of Transportation regulations.

  11. Senate examines measures to improve nuclear safety following Japan disaster

    NASA Astrophysics Data System (ADS)

    Showstack, Randy

    2012-03-01

    One year after Japan suffered a devastating magnitude 9.0 earthquake and the resulting tsunami and nuclear disaster, the U.S. Nuclear Regulatory Commission (NRC) has taken a number of measures to try to ensure that nuclear plants in the United States are safe from natural hazards. At a U.S. Senate hearing on 15 March, NRC chair Gregory Jaczko announced that the commission had issued three key orders and several requests for information on 12 March that plant licensees must follow, and that NRC also plans to take additional actions. However, the commission is not moving quickly enough in some areas, such as ensuring that all plants are safe from seismic hazards, including those in areas with low seismic activity, according to Jaczko's testimony before the Senate Committee on Environment and Public Works (EPW) and the Subcommittee on Clean Air and Nuclear Safety. The 12 March orders require licensees to have strategies to maintain or restore core cooling, containment, and spent-fuel pool cooling capabilities "following a beyond-design-basis extreme natural event" and have a reliable indication of the water level in spent-fuel storage pools.

  12. Nuclear safety. Technical progress journal, January--March 1996: Volume 37, No. 1

    SciTech Connect

    Muhlheim, M D

    1996-01-01

    Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but is also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Individual articles are indexed separately for the data base.

  13. Surveys of organizational culture and safety culture in nuclear power

    SciTech Connect

    Brown, Walter S.

    2000-07-30

    The results of a survey of organizational culture at a nuclear power plant are summarized and compared with those of a similar survey which has been described in the literature on ''high-reliability organizations''. A general-purpose cultural inventory showed a profile of organizational style similar to that reported in the literature; the factor structure for the styles was also similar to that of the plant previously described. A specialized scale designed to measure ''safety culture'' did not distinguished among groups within the organization that would be expected to differ.

  14. Status and Value of International Standards for Nuclear Criticality Safety

    SciTech Connect

    Hopper, Calvin Mitchell

    2011-01-01

    This presentation provides an update to the author's standards report provided at the ICNC-2007 meeting. It includes a discussion about the difference between, and the value of participating in, the development of international 'consensus' standards as opposed to nonconsensus standards. Standards are developed for a myriad of reasons. Generally, standards represent an agreed upon, repeatable way of doing something as defined by an individual or group of people. They come in various types. Examples include personal, family, business, industrial, commercial, and regulatory such as military, community, state, federal, and international standards. Typically, national and international 'consensus' standards are developed by individuals and organizations of diverse backgrounds representing the subject matter users and developers of a service or product and other interested parties or organizations. Within the International Organization for Standardization (ISO), Technical Committee 85 (TC85) on nuclear energy, Subcommittee 5 (SC5) on nuclear fuel technology, there is a Working Group 8 (WG8) on standardization of calculations, procedures, and practices related to criticality safety. WG8 has developed, and is developing, ISO standards within the category of nuclear criticality safety of fissionable materials outside of reactors (i.e., nonreactor fissionable material nuclear fuel cycle facilities). Since the presentation of the ICNC-2007 report, WG8 has issued three new finalized international standards and is developing two more new standards. Nearly all elements of the published WG8 ISO standards have been incorporated into IAEA nonconsensus guides and standards. The progression of consensus standards development among international partners in a collegial environment establishes a synergy of different concepts that broadens the perspectives of the members. This breadth of perspectives benefits the working group members in their considerations of consensus standards

  15. 76 FR 17460 - South Texas Project Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-29

    ... COMMISSION South Texas Project Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board..., 2.318, and 2.321, notice is hereby given that an Atomic Safety and Licensing Board (Board) is being...: Ronald M. Spritzer, Chair, Atomic Safety and Licensing Board Panel, U.S. Nuclear Regulatory...

  16. Worker Safety and Health and Nuclear Safety Quarterly Performance Analysis (January - March 2008)

    SciTech Connect

    Kerr, C E

    2009-10-07

    The DOE Office of Enforcement expects LLNL to 'implement comprehensive management and independent assessments that are effective in identifying deficiencies and broader problems in safety and security programs, as well as opportunities for continuous improvement within the organization' and to 'regularly perform assessments to evaluate implementation of the contractor's processes for screening and internal reporting.' LLNL has a self-assessment program, described in ES&H Manual Document 4.1, that includes line, management and independent assessments. LLNL also has in place a process to identify and report deficiencies of nuclear, worker safety and health and security requirements. In addition, the DOE Office of Enforcement expects LLNL to evaluate 'issues management databases to identify adverse trends, dominant problem areas, and potential repetitive events or conditions' (page 14, DOE Enforcement Process Overview, December 2007). LLNL requires that all worker safety and health and nuclear safety noncompliances be tracked as 'deficiencies' in the LLNL Issues Tracking System (ITS). Data from the ITS are analyzed for worker safety and health (WSH) and nuclear safety noncompliances that may meet the threshold for reporting to the DOE Noncompliance Tracking System (NTS). This report meets the expectations defined by the DOE Office of Enforcement to review the assessments conducted by LLNL, analyze the issues and noncompliances found in these assessments, and evaluate the data in the ITS database to identify adverse trends, dominant problem areas, and potential repetitive events or conditions. The report attempts to answer three questions: (1) Is LLNL evaluating its programs and state of compliance? (2) What is LLNL finding? (3) Is LLNL appropriately managing what it finds? The analysis in this report focuses on data from the first quarter of 2008 (January through March). This quarter is analyzed within the context of information identified in previous quarters to

  17. Renovated Korean nuclear safety and security system: A review and suggestions to successful settlement

    SciTech Connect

    Chung, W. S.; Yun, S. W.; Lee, D. S.; Go, D. Y.

    2012-07-01

    Questions of whether past nuclear regulatory body of Korea is not a proper system to monitor and check the country's nuclear energy policy and utilization have been raised. Moreover, a feeling of insecurity regarding nuclear safety after the nuclear accident in Japan has spread across the public. This has stimulated a renovation of the nuclear safety regime in Korea. The Nuclear Safety and Security Commission (NSSC) was launched on October 26, 2011 as a regulatory body directly under the President in charge of strengthening independence and nuclear safety. This was a meaningful event as the NSSC it is a much more independent regulatory system for Korea. However, the NSSC itself does not guarantee an enhanced public acceptance of the nuclear policy and stable use nuclear energy. This study introduces the new NSSC system and its details in terms of organization structure, appropriateness of specialty, budget stability, and management system. (authors)

  18. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 1: Reference Design Document (RDD)

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The Reference Design Document, of the Preliminary Safety Analysis Report (PSAR) - Reactor System provides the basic design and operations data used in the nuclear safety analysis of the Rector Power Module as applied to a Space Base program. A description of the power module systems, facilities, launch vehicle and mission operations, as defined in NASA Phase A Space Base studies is included. Each of two Zirconium Hydride Reactor Brayton power modules provides 50 kWe for the nominal 50 man Space Base. The INT-21 is the prime launch vehicle. Resupply to the 500 km orbit over the ten year mission is provided by the Space Shuttle. At the end of the power module lifetime (nominally five years), a reactor disposal system is deployed for boost into a 990 km high altitude (long decay time) earth orbit.

  19. Engineering thinking in emergency situations: A new nuclear safety concept.

    PubMed

    Guarnieri, Franck; Travadel, Sébastien

    2014-11-01

    The lessons learned from the Fukushima Daiichi accident have focused on preventive measures designed to protect nuclear reactors, and crisis management plans. Although there is still no end in sight to the accident that occurred on March 11, 2011, how engineers have handled the aftermath offers new insight into the capacity of organizations to adapt in situations that far exceed the scope of safety standards based on probabilistic risk assessment and on the comprehensive identification of disaster scenarios. Ongoing crises in which conventional resources are lacking, but societal expectations are high, call for "engineering thinking in emergency situations." This is a new concept that emphasizes adaptability and resilience within organizations-such as the ability to create temporary new organizational structures; to quickly switch from a normal state to an innovative mode; and to integrate a social dimension into engineering activities. In the future, nuclear safety oversight authorities should assess the ability of plant operators to create and implement effective engineering strategies on the fly, and should require that operators demonstrate the capability for resilience in the aftermath of an accident.

  20. Engineering thinking in emergency situations: A new nuclear safety concept

    PubMed Central

    Guarnieri, Franck; Travadel, Sébastien

    2014-01-01

    The lessons learned from the Fukushima Daiichi accident have focused on preventive measures designed to protect nuclear reactors, and crisis management plans. Although there is still no end in sight to the accident that occurred on March 11, 2011, how engineers have handled the aftermath offers new insight into the capacity of organizations to adapt in situations that far exceed the scope of safety standards based on probabilistic risk assessment and on the comprehensive identification of disaster scenarios. Ongoing crises in which conventional resources are lacking, but societal expectations are high, call for “engineering thinking in emergency situations.” This is a new concept that emphasizes adaptability and resilience within organizations—such as the ability to create temporary new organizational structures; to quickly switch from a normal state to an innovative mode; and to integrate a social dimension into engineering activities. In the future, nuclear safety oversight authorities should assess the ability of plant operators to create and implement effective engineering strategies on the fly, and should require that operators demonstrate the capability for resilience in the aftermath of an accident. PMID:25419015

  1. Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal

    NASA Astrophysics Data System (ADS)

    Herrero, J. J.; Rochman, D.; Leray, O.; Vasiliev, A.; Pecchia, M.; Ferroukhi, H.; Caruso, S.

    2017-09-01

    In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.

  2. Sensitivity-Uncertainty Based Nuclear Criticality Safety Validation

    SciTech Connect

    Brown, Forrest B.

    2016-09-20

    These are slides from a seminar given to the University of Mexico Nuclear Engineering Department. Whisper is a statistical analysis package developed to support nuclear criticality safety validation. It uses the sensitivity profile data for an application as computed by MCNP6 along with covariance files for the nuclear data to determine a baseline upper-subcritical-limit for the application. Whisper and its associated benchmark files are developed and maintained as part of MCNP6, and will be distributed with all future releases of MCNP6. Although sensitivity-uncertainty methods for NCS validation have been under development for 20 years, continuous-energy Monte Carlo codes such as MCNP could not determine the required adjoint-weighted tallies for sensitivity profiles. The recent introduction of the iterated fission probability method into MCNP led to the rapid development of sensitivity analysis capabilities for MCNP6 and the development of Whisper. Sensitivity-uncertainty based methods represent the future for NCS validation – making full use of today’s computer power to codify past approaches based largely on expert judgment. Validation results are defensible, auditable, and repeatable as needed with different assumptions and process models. The new methods can supplement, support, and extend traditional validation approaches.

  3. Safety and Nuclear Power Sources for Space Systems

    NASA Astrophysics Data System (ADS)

    Segalas, Corinne C.; Schmidt, George R.

    2010-09-01

    Nuclear power sources have been used in space applications for decades. They have been used extensively for electrical power production, and their future potential for propulsion has been recognized since the dawn of the spaceflight era. Nuclear power sources offer many advantages in terms of long duration operation and high power densities independent of distance and orientation with respect to the Sun. However, it is also broadly known that use of radioactive materials do carry more risk that must be addressed to ensure safe operation during all phases of the mission, particularly before and during launch into orbit. Almost all of the nuclear-powered missions to date have been flown by the United States and former Soviet Union, but other space-faring nations have recognized its importance for their future missions. Consequently, many in the space community have advocated the development of a broad set of principles that could be applied on an international basis. This paper examines the current guidelines by the major space-faring nations, and suggests a framework primarily based on the U.S. methodology for ensuring reduction of risk, mitigating environmental impact and promoting launch safety.

  4. Applications of Nuclear Data Covariances to Criticality Safety and Spent Fuel Characterization

    NASA Astrophysics Data System (ADS)

    Williams, M. L.; Ilas, G.; Marshall, W. J.; Rearden, B. T.

    2014-04-01

    Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

  5. Applications of nuclear data covariances to criticality safety and spent fuel characterization

    SciTech Connect

    Williams, Mark L; Ilas, Germina; Marshall, William BJ J; Rearden, Bradley T

    2014-01-01

    Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

  6. 10 CFR 1.42 - Office of Nuclear Material Safety and Safeguards.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... and secure production of nuclear fuel used in commercial nuclear reactors; the safe storage... nuclear materials, including certification of transport containers and reactor spent fuel storage; and... 10 Energy 1 2013-01-01 2013-01-01 false Office of Nuclear Material Safety and Safeguards....

  7. 10 CFR 1.42 - Office of Nuclear Material Safety and Safeguards.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... and secure production of nuclear fuel used in commercial nuclear reactors; the safe storage... nuclear materials, including certification of transport containers and reactor spent fuel storage; and... 10 Energy 1 2012-01-01 2012-01-01 false Office of Nuclear Material Safety and Safeguards....

  8. 10 CFR 1.42 - Office of Nuclear Material Safety and Safeguards.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... and secure production of nuclear fuel used in commercial nuclear reactors; the safe storage... nuclear materials, including certification of transport containers and reactor spent fuel storage; and... 10 Energy 1 2014-01-01 2014-01-01 false Office of Nuclear Material Safety and Safeguards....

  9. 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... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  10. 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... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  11. 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 license...

  12. 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... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

  13. 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 license...

  14. 78 FR 33449 - FirstEnergy Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-04

    ... COMMISSION FirstEnergy Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board Pursuant...: FirstEnergy Nuclear Operating Company This proceeding involves a license amendment request from FirstEnergy Nuclear Operating Company for Davis-Besse Nuclear Power Station, Unit 1, which is located...

  15. 76 FR 3678 - FirstEnergy Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-20

    ... COMMISSION FirstEnergy Nuclear Operating Company; Establishment of Atomic Safety and Licensing Board Pursuant... established to preside over the following proceeding: FirstEnergy Nuclear Operating Company (Davis-Besse Nuclear Power Station, Unit 1) This proceeding involves an application by FirstEnergy Nuclear...

  16. 78 FR 25488 - Qualification Tests for Safety-Related Actuators in Nuclear Power Plants

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-01

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment draft regulatory guide (DG), DG-1235, ``Qualification Tests for Safety-Related Actuators in Nuclear Power Plants.'' DG-1235 is proposed Revision 1 of RG 1.73, dated January 1974. This revision endorses, with clarifications, the enhanced consensus practices for qualifying safety-related actuators, and actuator components, in Nuclear Power Generating Stations in order to demonstrate their ability to perform their intended safety functions under all required conditions as described in the Institute of Electrical and Electronics Engineers (IEEE) standard 382-2006, ``Standard for Qualification of Safety-Related Actuators for Nuclear Power Generating Stations.''

  17. 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.

  18. Nuclear criticality safety evaluation of Spray Booth Operations in X-705, Portsmouth Gaseous Diffusion Plant

    SciTech Connect

    Sheaffer, M.K.; Keeton, S.C.

    1993-09-20

    This report evaluates nuclear criticality safety for Spray Booth Operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current procedures and related hardware/equipment is presented. Control parameters relevant to nuclear criticality safety are explained, and a consolidated listing of administrative controls and safety systems is developed. Based on compliance with DOE Orders and MMES practices, the overall operation is evaluated, and recommendations for enhanced safety are suggested.

  19. Implementing 10 CFR 830 at the FEMP Silos: Nuclear Health and Safety Plans as Documented Safety Analysis

    SciTech Connect

    Fisk, Patricia; Rutherford, Lavon

    2003-06-01

    The objective of the Silos Project at the Fernald Closure Project (FCP) is to safely remediate high-grade uranium ore residues (Silos 1 and 2) and metal oxide residues (Silo 3). The evolution of Documented Safety Analyses (DSAs) for these facilities has reflected the changes in remediation processes. The final stage in silos DSAs is an interpretation of 10 CFR 830 Safe Harbor Requirements that combines a Health and Safety Plan with nuclear safety requirements. This paper will address the development of a Nuclear Health and Safety Plan, or N-HASP.

  20. 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).

  1. 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.

  2. Improved nuclear power plant operations and safety through performance-based safety regulation.

    PubMed

    Golay, M W

    2000-01-07

    This paper illustrates some of the promise and needed future work for risk-informed, performance-based regulation (RIPBR). RIPBR is an evolving alternative to the current prescriptive method of nuclear safety regulation. Prescriptive regulation effectively constitutes a long, fragmented checklist of requirements that safety-related systems in a plant must satisfy. RIPBR, instead, concentrates upon satisfying negotiated performance goals and incentives for judging and rewarding licensee behavior to improve safety and reduce costs. In a project reported here, a case study was conducted concerning a pressurized water reactor (PWR) emergency diesel generator (EDG). Overall, this work has shown that the methods of RIPBR are feasible to use, and capable of justifying simultaneous safety and economic nuclear power improvements. However, it also reveals several areas where the framework of RIPBR should be strengthened. First, researchers need better data and understanding regarding individual component-failure modes that may cause components to fail. Not only are more data needed on failure rates, but more data and understanding are needed to enable analysts to evaluate whether these failures become more likely as the interval between tests is increased. This is because the current state of failure data is not sufficiently finely detailed to define the failure rates of individual component failure modes; such knowledge is needed when changing component-specific regulatory requirements. Second, the role of component testing, given that a component has failed, needs to be strengthened within the context of RIPBR. This includes formulating requirements for updating the prior probability distribution of a component failure rate and conducting additional or more frequent testing. Finally, as a means of compensating for unavoidable uncertainty as an obstacle to regulatory decision-making, limits to knowledge must be treated explicitly and formally. This treatment includes the

  3. Feedback from uncertainties propagation research projects conducted in different hydraulic fields: outcomes for engineering projects and nuclear safety assessment.

    NASA Astrophysics Data System (ADS)

    Bacchi, Vito; Duluc, Claire-Marie; Bertrand, Nathalie; Bardet, Lise

    2017-04-01

    In recent years, in the context of hydraulic risk assessment, much effort has been put into the development of sophisticated numerical model systems able reproducing surface flow field. These numerical models are based on a deterministic approach and the results are presented in terms of measurable quantities (water depths, flow velocities, etc…). However, the modelling of surface flows involves numerous uncertainties associated both to the numerical structure of the model, to the knowledge of the physical parameters which force the system and to the randomness inherent to natural phenomena. As a consequence, dealing with uncertainties can be a difficult task for both modelers and decision-makers [Ioss, 2011]. In the context of nuclear safety, IRSN assesses studies conducted by operators for different reference flood situations (local rain, small or large watershed flooding, sea levels, etc…), that are defined in the guide ASN N°13 [ASN, 2013]. The guide provides some recommendations to deal with uncertainties, by proposing a specific conservative approach to cover hydraulic modelling uncertainties. Depending of the situation, the influencing parameter might be the Strickler coefficient, levee behavior, simplified topographic assumptions, etc. Obviously, identifying the most influencing parameter and giving it a penalizing value is challenging and usually questionable. In this context, IRSN conducted cooperative (Compagnie Nationale du Rhone, I-CiTy laboratory of Polytech'Nice, Atomic Energy Commission, Bureau de Recherches Géologiques et Minières) research activities since 2011 in order to investigate feasibility and benefits of Uncertainties Analysis (UA) and Global Sensitivity Analysis (GSA) when applied to hydraulic modelling. A specific methodology was tested by using the computational environment Promethee, developed by IRSN, which allows carrying out uncertainties propagation study. This methodology was applied with various numerical models and in

  4. Enforcement handbook: Enforcement of DOE nuclear safety requirements

    SciTech Connect

    1995-06-01

    This Handbook provides detailed guidance and procedures to implement the General Statement of DOE Enforcement Policy (Enforcement Policy or Policy). A copy of this Enforcement Policy is included for ready reference in Appendix D. The guidance provided in this Handbook is qualified, however, by the admonishment to exercise discretion in determining the proper disposition of each potential enforcement action. As discussed in subsequent chapters, the Enforcement and Investigation Staff will apply a number of factors in assessing each potential enforcement situation. Enforcement sanctions are imposed in accordance with the Enforcement Policy for the purpose of promoting public and worker health and safety in the performance of activities at DOE facilities by DOE contractors (and their subcontractors and suppliers) who are indemnified under the Price-Anderson Amendments Act. These indemnified contractors, and their suppliers and subcontractors, will be referred to in this Handbook collectively as DOE contractors. It should be remembered that the purpose of the Department`s enforcement policy is to improve nuclear safety for the workers and the public, and this goal should be the prime consideration in exercising enforcement discretion.

  5. 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.

  6. Importance of bladder radioactivity for radiation safety in nuclear medicine.

    PubMed

    Gültekin, Salih Sinan; Sahmaran, Turan

    2013-12-01

    Most of the radiopharmaceuticals used in nuclear medicine are excreted via the urinary system. This study evaluated the importance of a reduction in bladder radioactivity for radiation safety. The study group of 135 patients underwent several organ scintigraphies [40/135; thyroid scintigraphy (TS), 30/135; whole body bone scintigraphy (WBS), 35/135; myocardial perfusion scintigraphy (MPS) and 30/135; renal scintigraphy (RS)] by a technologist within 1 month. In full and empty conditions, static bladder images and external dose rate measurements at 0.25, 0.50, 1, 1.5 and 2 m distances were obtained and decline ratios were calculated from these two data sets. External radiation dose rates were highest in patients undergoing MPS. External dose rates at 0.25 m distance for TS, TKS, MPS and BS were measured to be 56, 106, 191 and 72 μSv h-1 for full bladder and 29, 55, 103 and 37 μSv h-1 for empty bladder, respectively. For TS, WBS, MPS and RS, respectively, average decline ratios were calculated to be 52%, 55%, 53% and 54% in the scintigraphic assessment and 49%, 51%, 49%, 50% and 50% in the assessment with Geiger counter. Decline in bladder radioactivity is important in terms of radiation safety. Patients should be encouraged for micturition after each scintigraphic test. Spending time together with radioactive patients at distances less than 1 m should be kept to a minimum where possible. None declared.

  7. Nuclear Safety Analysis for the Mars Exploration Rover 2003 Project

    NASA Astrophysics Data System (ADS)

    Firstenberg, Henry; Rutger, Lyle L.; Mukunda, Meera; Bartram, Bart W.

    2004-02-01

    The National Aeronautics and Space Administration's Mars Exploration Rover (MER) 2003 project is designed to place two mobile laboratories (Rovers) on Mars to remotely characterize a diversity of rocks and soils. Milestones accomplished so far include two successful launches of identical spacecraft (the MER-A and MER-B missions) from Cape Canaveral Air Force Station, Florida on June 10 and July 7, 2003. Each Rover uses eight Light Weight Radioisotope Heater Units (LWRHUs) fueled with plutonium-238 dioxide to provide local heating of Rover components. The LWRHUs are provided by the U.S. Department of Energy. In addition, small quantities of radioactive materials in sealed sources are used in scientific instrumentation on the Rover. Due to the radioactive nature of these materials and the potential for accidents, a formal Launch Approval Process requires the preparation of a Final Safety Analysis Report (FSAR) for submittal to and independent review by an Interagency Nuclear Safety Review Panel. This paper presents a summary of the FSAR in terms of potential accident scenarios, probabilities, source terms, radiological consequences, mission risks, and uncertainties in the reported results.

  8. Fuzzy-logic-based safety verification framework for nuclear power plants.

    PubMed

    Rastogi, Achint; Gabbar, Hossam A

    2013-06-01

    This article presents a practical implementation of a safety verification framework for nuclear power plants (NPPs) based on fuzzy logic where hazard scenarios are identified in view of safety and control limits in different plant process values. Risk is estimated quantitatively and compared with safety limits in real time so that safety verification can be achieved. Fuzzy logic is used to define safety rules that map hazard condition with required safety protection in view of risk estimate. Case studies are analyzed from NPP to realize the proposed real-time safety verification framework. An automated system is developed to demonstrate the safety limit for different hazard scenarios. © 2012 Society for Risk Analysis.

  9. Assessment of the safety of US nuclear weapons and related nuclear test requirements: A post-Bush Initiative update

    SciTech Connect

    Kidder, R.E.

    1991-12-10

    The Nuclear Weapons Reduction Initiative announced by President Bush on September 27, 1991, is described herein as set forth in Defense Secretary Cheney`s Nuclear Arsenal Reduction Order issued September 28, 1991. The implications of the Bush Initiative for improved nuclear weapons safety are assessed in response to a request by US Senators Harkin, Kennedy, and Wirth to the Lawrence Livermore National Laboratory that the author prepare such an assessment. The author provides an estimate of the number of nuclear tests needed to accomplish a variety of specified warhead safety upgrades, then uses the results of this estimate to answer three questions posed by the Senators. These questions concern pit reuse and the number of nuclear tests needed for specified safety upgrades of those ballistic missiles not scheduled for retirement, namely the Minuteman III, C4, and D5 missiles.

  10. Manned space flight nuclear system safety. Volume 3: Reactor system preliminary nuclear safety analysis. Part 2: Accident Model Document (AMD)

    NASA Technical Reports Server (NTRS)

    1972-01-01

    The Accident Model Document is one of three documents of the Preliminary Safety Analysis Report (PSAR) - Reactor System as applied to a Space Base Program. Potential terrestrial nuclear hazards involving the zirconium hydride reactor-Brayton power module are identified for all phases of the Space Base program. The accidents/events that give rise to the hazards are defined and abort sequence trees are developed to determine the sequence of events leading to the hazard and the associated probabilities of occurence. Source terms are calculated to determine the magnitude of the hazards. The above data is used in the mission accident analysis to determine the most probable and significant accidents/events in each mission phase. The only significant hazards during the prelaunch and launch ascent phases of the mission are those which arise form criticality accidents. Fission product inventories during this time period were found to be very low due to very limited low power acceptance testing.

  11. 33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

    Code of Federal Regulations, 2014 CFR

    2014-07-01

    ... 33 Navigation and Navigable Waters 2 2014-07-01 2014-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

  12. 33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

    Code of Federal Regulations, 2013 CFR

    2013-07-01

    ... 33 Navigation and Navigable Waters 2 2013-07-01 2013-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

  13. 33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

    Code of Federal Regulations, 2010 CFR

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

  14. 33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

    Code of Federal Regulations, 2012 CFR

    2012-07-01

    ... 33 Navigation and Navigable Waters 2 2012-07-01 2012-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

  15. 33 CFR 165.115 - Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts.

    Code of Federal Regulations, 2011 CFR

    2011-07-01

    ... 33 Navigation and Navigable Waters 2 2011-07-01 2011-07-01 false Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth, Massachusetts. 165.115 Section 165.115 Navigation and Navigable... Coast Guard District § 165.115 Safety and Security Zones; Pilgrim Nuclear Power Plant, Plymouth...

  16. SRTC criticality safety technical review: Nuclear Criticality Safety Evaluation 93-04 enriched uranium receipt

    SciTech Connect

    Rathbun, R.

    1993-10-13

    Review of NMP-NCS-930087, {open_quotes}Nuclear Criticality Safety Evaluation 93-04 Enriched Uranium Receipt (U), July 30, 1993, {close_quotes} was requested of SRTC (Savannah River Technology Center) Applied Physics Group. The NCSE is a criticality assessment to determine the mass limit for Engineered Low Level Trench (ELLT) waste uranium burial. The intent is to bury uranium in pits that would be separated by a specified amount of undisturbed soil. The scope of the technical review, documented in this report, consisted of (1) an independent check of the methods and models employed, (2) independent HRXN/KENO-V.a calculations of alternate configurations, (3) application of ANSI/ANS 8.1, and (4) verification of WSRC Nuclear Criticality Safety Manual procedures. The NCSE under review concludes that a 500 gram limit per burial position is acceptable to ensure the burial site remains in a critically safe configuration for all normal and single credible abnormal conditions. This reviewer agrees with that conclusion.

  17. Style, content and format guide for writing safety analysis documents. Volume 1, Safety analysis reports for DOE nuclear facilities

    SciTech Connect

    Not Available

    1994-06-01

    The purpose of Volume 1 of this 4-volume style guide is to furnish guidelines on writing and publishing Safety Analysis Reports (SARs) for DOE nuclear facilities at Sandia National Laboratories. The scope of Volume 1 encompasses not only the general guidelines for writing and publishing, but also the prescribed topics/appendices contents along with examples from typical SARs for DOE nuclear facilities.

  18. Nuclear Safety. Technical Progress Journal, April--June 1993: Volume 34, No. 2

    SciTech Connect

    Silver, E G

    1993-01-01

    This review journal that covers significant developments in the field of Nuclear Safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations are also treated.

  19. Nuclear Safety. Technical Progress Journal, October--December 1991: Volume 32, No. 4

    SciTech Connect

    Not Available

    1991-01-01

    This document is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  20. Nuclear Safety Technical Progress Journal, January--June 1995. Volume 36, No. 1

    SciTech Connect

    Silver, E G

    1995-01-01

    This journal covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  1. Nuclear Safety. Technical Progress Journal, July--September 1992: Volume 33, No. 3

    SciTech Connect

    1992-01-01

    This review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated.

  2. 76 FR 42686 - DOE Response to Recommendation 2011-1 of the Defense Nuclear Facilities Safety Board, Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-19

    ... workers, the public, and the environment. DOE line management is both responsible and accountable for.... We hold our contractors to the same standard. A strong nuclear safety and quality culture is the... Environmental Management (EM) requested that HSS conduct a comprehensive analysis of the safety culture at...

  3. Optimization of a Dry, Mixed Nuclear Fuel Storage Array for Nuclear Criticality Safety

    NASA Astrophysics Data System (ADS)

    Baranko, Benjamin T.

    A dry storage array of used nuclear fuel at the Idaho National Laboratory contains a mixture of more than twenty different research and test reactor fuel types in up to 636 fuel storage canisters. New analysis demonstrates that the current arrangement of the different fuel-type canisters does not minimize the system neutron multiplication factor (keff), and that the entire facility storage capacity cannot be utilized without exceeding the subcritical limit (ksafe) for ensuring nuclear criticality safety. This work determines a more optimal arrangement of the stored fuels with a goal to minimize the system keff, but with a minimum of potential fuel canister relocation movements. The solution to this multiple-objective optimization problem will allow for both an improvement in the facility utilization while also offering an enhancement in the safety margin. The solution method applies stochastic approximation and a Tabu search metaheuristic to an empirical model developed from supporting MCNP calculations. The results establish an optimal relocation of between four to sixty canisters, which will allow the current thirty-one empty canisters to be used for storage while reducing the array keff by up to 0.018 +/- 0.003 relative to the current arrangement.

  4. Organizational Culture for Safety, Security, and Safeguards in New Nuclear Power Countries

    SciTech Connect

    Kovacic, Donald N

    2015-01-01

    This chapter will contain the following sections: Existing international norms and standards for developing the infrastructure to support new nuclear power programs The role of organizational culture and how it supports the safe, secure, and peaceful application of nuclear power Identifying effective and efficient strategies for implementing safety, security and safeguards in nuclear operations Challenges identified in the implementation of safety, security and safeguards Potential areas for future collaboration between countries in order to support nonproliferation culture

  5. Proceedings of the international meeting on thermal nuclear reactor safety. Vol. 1

    SciTech Connect

    1983-02-01

    Separate abstracts are included for each of the papers presented concerning current issues in nuclear power plant safety; national programs in nuclear power plant safety; radiological source terms; probabilistic risk assessment methods and techniques; non LOCA and small-break-LOCA transients; safety goals; pressurized thermal shocks; applications of reliability and risk methods to probabilistic risk assessment; human factors and man-machine interface; and data bases and special applications.

  6. Nuclear Safety Functions of ITER Gas Injection System Instrumentation and Control and the Concept Design

    NASA Astrophysics Data System (ADS)

    Yang, Yu; Maruyama, S.; Fossen, A.; Villers, F.; Kiss, G.; Zhang, Bo; Li, Bo; Jiang, Tao; Huang, Xiangmei

    2016-08-01

    The ITER Gas Injection System (GIS) plays an important role on fueling, wall conditioning and distribution for plasma operation. Besides that, to support the safety function of ITER, GIS needs to implement three nuclear safety Instrumentation and Control (I&C) functions. In this paper, these three functions are introduced with the emphasis on their latest safety classifications. The nuclear I&C design concept is briefly discussed at the end.

  7. Real-time graphic display utility for nuclear safety applications

    SciTech Connect

    Yang, S.; Huang, X.; Taylor, J.; Stevens, J.; Gerardis, T.; Hsu, A.; McCreary, T.

    2006-07-01

    With the increasing interests in the nuclear energy, new nuclear power plants will be constructed and licensed, and older generation ones will be upgraded for assuring continuing operation. The tendency of adopting the latest proven technology and the fact of older parts becoming obsolete have made the upgrades imperative. One of the areas for upgrades is the older CRT display being replaced by the latest graphics displays running under modern real time operating system (RTOS) with safety graded modern computer. HFC has developed a graphic display utility (GDU) under the QNX RTOS. A standard off-the-shelf software with a long history of performance in industrial applications, QNX RTOS used for safety applications has been examined via a commercial dedication process that is consistent with the regulatory guidelines. Through a commercial survey, a design life cycle and an operating history evaluation, and necessary tests dictated by the dedication plan, it is reasonably confirmed that the QNX RTOS was essentially equivalent to what would be expected in the nuclear industry. The developed GDU operates and communicates with the existing equipment through a dedicated serial channel of a flat panel controller (FPC) module. The FPC module drives a flat panel display (FPD) monitor. A touch screen mounted on the FPD serves as the normal operator interface with the FPC/FPD monitor system. The GDU can be used not only for replacing older CRTs but also in new applications. The replacement of the older CRT does not disturb the function of the existing equipment. It not only provides modern proven technology upgrade but also improves human ergonomics. The FPC, which can be used as a standalone controller running with the GDU, is an integrated hardware and software module. It operates as a single board computer within a control system, and applies primarily to the graphics display, targeting, keyboard and mouse. During normal system operation, the GDU has two sources of data

  8. The Gulf Nuclear Energy Infrastructure Institute : an integrated approach to safety, security & safeguards.

    SciTech Connect

    Williams, Adam David

    2010-04-01

    Sandia National Laboratories (SNL) and the Nuclear Security Science and Policy Institute (NSSPI) at Texas A&M University are working with Middle East regional partners to set up a nuclear energy safety, safeguards, and security educational institute in the Gulf region. SNL and NSSPI, partnered with the Khalifa University of Science, Technology, and Research (KUSTAR), with suppot from its key nuclear stakeholders, the Emirates Nuclear Energy Corporation (ENEC), and the Federal Authority for Nuclear Regulation (FANR), plan to jointly establish the institute in Abu Dhabi. The Gulf Nuclear Energy Infrastructure Institute (GNEII) will be a KUSTAR-associated, credit-granting regional education program providing both classroom instruction and hands-on experience. The ultimate objective is for GNEII to be autonomous - regionally funded and staffed with personnel capable of teaching all GNEII courses five years after its inauguration. This is a strategic effort to indigenize a responsible nuclear energy culture - a culture shaped by an integrated understanding of nuclear safety, safeguards and security - in regional nuclear energy programs. GNEII also promotes international interests in developing a nuclear energy security and safety culture, increases collaboration between the nuclear energy security and safety communities, and helps to enhance global standards for nuclear energy technology in the Middle East.

  9. The Gulf Nuclear Energy Infrastructure Institute : an integrated approach to safety, security and safeguards.

    SciTech Connect

    Beeley, Phillip A.; Boyle, David R.; Williams, Adam David; Mohagheghi, Amir Hossein

    2010-04-01

    Sandia National Laboratories (SNL) and the Nuclear Security Science and Policy Institute (NSSPI) at Texas A&M University are working with Middle East regional partners to set up a nuclear energy safety, safeguards, and security educational institute in the Gulf region. SNL and NSSPI, partnered with the Khalifa University of Science, Technology, and Research (KUSTAR), with suppot from its key nuclear stakeholders, the Emirates Nuclear Energy Corporation (ENEC), and the Federal Authority for Nuclear Regulation (FANR), plan to jointly establish the institute in Abu Dhabi. The Gulf Nuclear Energy Infrastructure Institute (GNEII) will be a KUSTAR-associated, credit-granting regional education program providing both classroom instruction and hands-on experience. The ultimate objective is for GNEII to be autonomous - regionally funded and staffed with personnel capable of teaching all GNEII courses five years after its inauguration. This is a strategic effort to indigenize a responsible nuclear energy culture - a culture shaped by an integrated understanding of nuclear safety, safeguards and security - in regional nuclear energy programs. GNEII also promotes international interests in developing a nuclear energy security and safety culture, increases collaboration between the nuclear energy security and safety communities, and helps to enhance global standards for nuclear energy technology in the Middle East.

  10. Extreme Storm Event Assessments for Nuclear Facilities and Dam Safety

    NASA Astrophysics Data System (ADS)

    England, J. F.; Nicholson, T. J.; Prasad, R.

    2008-12-01

    Extreme storm events over the last 35 years are being assessed to evaluate flood estimates for safety assessments of dams, nuclear power plants, and other high-hazard structures in the U.S. The current storm rainfall design standard for evaluating the flood potential at dams and non-coastal nuclear power plants is the Probable Maximum Precipitation (PMP). PMP methods and estimates are published in the National Weather Service generalized hydrometeorological reports (HMRs). A new Federal Interagency cooperative effort is reviewing hydrometeorologic data from large storms which have occurred in the last 20 to 40 years and were not included in the database used in the development of many of the HMRs. Extreme storm data, such as the January 1996 storm in Pennsylvania, June 2008 Iowa storms, and Hurricanes Andrew (1992), Floyd (1999), Isabel (2003), Katrina (2005), need to be systematically assembled and analyzed for use in these regional extreme storm studies. Storm maximization, transposition, envelopment, and depth-area duration procedures will incorporate recent advances in hydrometeorology, including radar precipitation data and stochastic storm techniques. We describe new cooperative efforts to develop a database of extreme storms and to examine the potential impacts of recent extreme storms on PMP estimates. These efforts will be coordinated with Federal agencies, universities, and the private sector through an Extreme Storm Events Work Group under the Federal Subcommittee on Hydrology. This work group is chartered to coordinate studies and develop databases for reviewing and improving methodologies and data collection techniques used to estimate design precipitation up to and including the PMP. The initial effort focuses on collecting and reviewing extreme storm event data in the Southeastern U.S. that have occurred since Tropical Storm Agnes (1972). Uncertainties and exceedance probability estimates of PMP are being explored. Potential effects of climate

  11. Importance of Bladder Radioactivity for Radiation Safety in Nuclear Medicine

    PubMed Central

    Gültekin, Salih Sinan; Şahmaran, Turan

    2013-01-01

    Objective: Most of the radiopharmaceuticals used in nuclear medicine are excreted via the urinary system. This study evaluated the importance of a reduction in bladder radioactivity for radiation safety. Methods: The study group of 135 patients underwent several organ scintigraphies [40/135; thyroid scintigraphy (TS), 30/135; whole body bone scintigraphy (WBS), 35/135; myocardial perfusion scintigraphy (MPS) and 30/135; renal scintigraphy (RS)] by a technologist within 1 month. In full and empty conditions, static bladder images and external dose rate measurements at 0.25, 0.50, 1, 1.5 and 2 m distances were obtained and decline ratios were calculated from these two data sets. Results: External radiation dose rates were highest in patients undergoing MPS. External dose rates at 0.25 m distance for TS, TKS, MPS and BS were measured to be 56, 106, 191 and 72 μSv h-1 for full bladder and 29, 55, 103 and 37 μSv h-1 for empty bladder, respectively. For TS, WBS, MPS and RS, respectively, average decline ratios were calculated to be 52%, 55%, 53% and 54% in the scintigraphic assessment and 49%, 51%, 49%, 50% and 50% in the assessment with Geiger counter. Conclusion: Decline in bladder radioactivity is important in terms of radiation safety. Patients should be encouraged for micturition after each scintigraphic test. Spending time together with radioactive patients at distances less than 1 m should be kept to a minimum where possible. Conflict of interest:None declared. PMID:24416625

  12. 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.

  13. Nuclear criticality safety modeling of an LEU deposit

    SciTech Connect

    Haire, M.J.; Elam, K.R.; Jordan, W.C.; Dahl, T.L.

    1996-11-01

    The construction of the Oak Ridge Gaseous Diffusion Plant (now known as the K-25 Site) began during World War H and eventually consisted of five major process buildings: K-25, K-27, K-29, K-31, and K-33. The plant took natural (0.711% {sup 231}U) uranium as feed and processed it into both low-enriched uranium (LEU) and high-enriched uranium (HEU) with concentrations up to {approximately}93% {sup 231}U. The K-25 and K-27 buildings were shut down in 1964, but the rest of the plant produced LEU until 1985. During operation, inleakage of humid air into process piping and equipment caused reactions with gaseous uranium hexafluoride (UF{sub 6}) that produced nonvolatile uranyl fluoride (UO{sub 2}F{sub 2}) deposits. As part of shutdown, most of the uranium was evacuated as volatile UF{sub 6}. The UO{sub 2}F{sub 2} deposits remained. The U.S. Department of Energy has mitiated a program to unprove nuclear criticality safety by removing the larger enriched uranium deposits.

  14. Nuclear criticality safety program for environmental restoration projects

    SciTech Connect

    Marble, R.C.; Brown, T.D.

    1994-05-01

    The Fernald Environmental Management Project (FEMP), formerly known as the Feed Materials Production Center (FMPC), is located on a 1050 acre site approximately twenty miles northwest of Cincinnati, Ohio. The production area of the site covers approximately 136 acres in the central portion of the site. Surrounding the core production area is a buffer consisting of leased grazing land, reforested land, and unused areas. The uranium processing facility was designed and constructed in the early 1950s. During the period from 1952 to 1989 the site produced uranium feed material and uranium products used in the United States weapons complex. Production at the site ended in 1989, when the site was shut down for what was expected to be a short period of time. However, the FUTC was permanently shut down in 1991, and the site`s mission was changed from production to environmental restoration. The objective of this paper is to give an update on activities at the Fernald Site and to describe the Nuclear Criticality Safety issues that are currently being addressed.

  15. Linking empowering leadership to safety participation in nuclear power plants: a structural equation model.

    PubMed

    Martínez-Córcoles, Mario; Schöbel, Markus; Gracia, Francisco J; Tomás, Inés; Peiró, José M

    2012-07-01

    Safety participation is of paramount importance in guaranteeing the safe running of nuclear power plants. The present study examined the effects of empowering leadership on safety participation. Based on a sample of 495 employees from two Spanish nuclear power plants, structural equation modeling showed that empowering leadership has a significant relationship with safety participation, which is mediated by collaborative team learning. In addition, the results revealed that the relationship between empowering leadership and collaborative learning is partially mediated by the promotion of dialogue and open communication. The implications of these findings for safety research and their practical applications are outlined. An empowering leadership style enhances workers' safety performance, particularly safety participation behaviors. Safety participation is recommended to detect possible rule inconsistencies or misunderstood procedures and make workers aware of critical safety information and issues. Crown Copyright © 2012. Published by Elsevier Ltd. All rights reserved.

  16. 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)

  17. Training and qualification program for nuclear criticality safety technical staff. Revision 1

    SciTech Connect

    Taylor, R.G.; Worley, C.A.

    1997-03-05

    A training and qualification program for nuclear criticality safety technical staff personnel has been developed and implemented. All personnel who are to perform nuclear criticality safety technical work are required to participate in the program. The program includes both general nuclear criticality safety and plant specific knowledge components. Advantage can be taken of previous experience for that knowledge which is portable such as performance of computer calculations. Candidates step through a structured process which exposes them to basic background information, general plant information, and plant specific information which they need to safely and competently perform their jobs. Extensive documentation is generated to demonstrate that candidates have met the standards established for qualification.

  18. Border safety: quality control at the nuclear envelope

    PubMed Central

    Webster, Brant M.; Lusk, C. Patrick

    2015-01-01

    The unique biochemical identity of the nuclear envelope confers its capacity to establish a barrier that protects the nuclear compartment and directly contributes to nuclear function. Recent work uncovered quality control mechanisms employing the ESCRT machinery and a new arm of ERAD to counteract the unfolding, damage or misassembly of nuclear envelope proteins and ensure the integrity of the nuclear envelope membranes. Moreover, cells have the capacity to recognize and triage defective nuclear pore complexes in order to prevent their inheritance and preserve the longevity of progeny. These mechanisms serve to highlight the diverse strategies used by cells to maintain nuclear compartmentalization; we suggest they mitigate the progression and severity of diseases associated with nuclear envelope malfunction like the laminopathies. PMID:26437591

  19. Nuclear criticality safety aspects of gaseous uranium hexafluoride (UF{sub 6}) in the diffusion cascade

    SciTech Connect

    Huffer, J.E.

    1997-04-01

    This paper determines the nuclear safety of gaseous UF{sub 6} in the current Gaseous Diffusion Cascade and auxiliary systems. The actual plant safety system settings for pressure trip points are used to determine the maximum amount of HF moderation in the process gas, as well as the corresponding atomic number densities. These inputs are used in KENO V.a criticality safety models which are sized to the actual plant equipment. The ENO V.a calculation results confirm nuclear safety of gaseous UF{sub 6} in plant operations..

  20. Nuclear criticality safety calculational analysis for small-diameter containers

    SciTech Connect

    LeTellier, M.S.; Smallwood, D.J.; Henkel, J.A.

    1995-11-01

    This report documents calculations performed to establish a technical basis for the nuclear criticality safety of favorable geometry containers, sometimes referred to as 5-inch containers, in use at the Portsmouth Gaseous Diffusion Plant. A list of containers currently used in the plant is shown in Table 1.0-1. These containers are currently used throughout the plant with no mass limits. The use of containers with geometries or material types other than those addressed in this evaluation must be bounded by this analysis or have an additional analysis performed. The following five basic container geometries were modeled and bound all container geometries in Table 1.0-1: (1) 4.32-inch-diameter by 50-inch-high polyethylene bottle; (2) 5.0-inch-diameter by 24-inch-high polyethylene bottle; (3) 5.25-inch-diameter by 24-inch-high steel can ({open_quotes}F-can{close_quotes}); (4) 5.25-inch-diameter by 15-inch-high steel can ({open_quotes}Z-can{close_quotes}); and (5) 5.0-inch-diameter by 9-inch-high polybottle ({open_quotes}CO-4{close_quotes}). Each container type is evaluated using five basic reflection and interaction models that include single containers and multiple containers in normal and in credible abnormal conditions. The uranium materials evaluated are UO{sub 2}F{sub 2}+H{sub 2}O and UF{sub 4}+oil materials at 100% and 10% enrichments and U{sub 3}O{sub 8}, and H{sub 2}O at 100% enrichment. The design basis safe criticality limit for the Portsmouth facility is k{sub eff} + 2{sigma} < 0.95. The KENO study results may be used as the basis for evaluating general use of these containers in the plant.

  1. Status report of the US Department of Energy`s International Nuclear Safety Program

    SciTech Connect

    1994-12-01

    The US Department of Energy (DOE) implements the US Government`s International Nuclear Safety Program to improve the level of safety at Soviet-designed nuclear power plants in Central and Eastern Europe, Russia, and Unkraine. The program is conducted consistent with guidance and policies established by the US Department of State (DOS) and the Agency for International Development and in close collaboration with the Nuclear Regulatory Commission. Some of the program elements were initiated in 1990 under a bilateral agreement with the former Soviet Union; however, most activities began after the Lisbon Nuclear Safety Initiative was announced by the DOS in 1992. Within DOE, the program is managed by the International Division of the Office of Nuclear Energy. The overall objective of the International Nuclear Safety Program is to make comprehensive improvements in the physical conditions of the power plants, plant operations, infrastructures, and safety cultures of countries operating Soviet-designed reactors. This status report summarizes the Internatioal Nuclear Safety Program`s activities that have been completed as of September 1994 and discusses those activities currently in progress.

  2. Safety and environmental analyses for space nuclear programs

    NASA Technical Reports Server (NTRS)

    Mcconnell, D. J.

    1990-01-01

    The tools and procedures for analyzing environmental quality and safety are reviewed. The process of preparing an environmental impact statement is outlined and the data sources for a safety analysis are discussed. The environmental safety analysis process is demonstrated, using examples from the Galileo, Ulysses, and Venus-earth-earth-gravity-assist programs.

  3. Nuclear energy with inherent safety: Change of outdated paradigm, criteria

    NASA Astrophysics Data System (ADS)

    Adamov, E. O.; Orlov, V. V.; Rachkov, V. I.; Slessarev, I. S.; Khomyakov, Yu. S.

    2015-12-01

    Modern nuclear power technology still has significant sources of risk, and, weak links, such as, a threat of severe accidents with catastrophic unpredictable consequences and damage to the population, proliferation of nuclear weapon-usable materials, risks of long-term storage of toxic radioactive waste, risks of loss of major investments in nuclear facilities and their construction, lack of fuel resources for the ambitious role of nuclear power in the competitive balance of energy. Each of these risks is important and almost independent, though the elimination of some of them does not significantly alter the overall assessment of nuclear power.

  4. Preparation, review, and approval of implementation plans for nuclear safety requirements

    SciTech Connect

    Not Available

    1994-10-01

    This standard describes an acceptable method to prepare, review, and approve implementation plans for DOE Nuclear Safety requirements. DOE requirements are identified in DOE Rules, Orders, Notices, Immediate Action Directives, and Manuals.

  5. Manual of functions, assignments, and responsibilities for nuclear safety: Revision 2

    SciTech Connect

    Not Available

    1994-10-15

    The FAR Manual is a convenient easy-to-use collection of the functions, assignments, and responsibilities (FARs) of DOE nuclear safety personnel. Current DOE directives, including Orders, Secretary of Energy Notices, and other assorted policy memoranda, are the source of this information and form the basis of the FAR Manual. Today, the majority of FARs for DOE personnel are contained in DOE`s nuclear safety Orders. As these Orders are converted to rules in the Code of Federal Regulations, the FAR Manual will become the sole source for information relating to the functions, assignments, responsibilities of DOE nuclear safety personnel. The FAR Manual identifies DOE directives that relate to nuclear safety and the specific DOE personnel who are responsible for implementing them. The manual includes only FARs that have been extracted from active directives that have been approved in accordance with the procedures contained in DOE Order 1321.1B.

  6. Synthesis of the C-terminal octapeptide of pig oxyntomodulin. Lys-Arg-Asn-Lys-Asn-Asn-Ile-Ala: a potent inhibitor of pentagastrin-induced acid secretion.

    PubMed

    Audousset-Puech, M P; Jarrousse, C; Dubrasquet, M; Aumelas, A; Castro, B; Bataille, D; Martinez, J

    1985-10-01

    The synthesis of Lys-Arg-Asn-Lys-Asn-Asn-Ile-Ala representing the C-terminal octapeptide of oxyntomodulin isolated from pig intestine is described. Its structure was confirmed by its 360-MHz 1H NMR spectra. The octapeptide was tested for its ability to inhibit pentagastrin-induced acid secretion, in the anaesthetized rat, in the conscious rat with chronic gastric fistula, and in the conscious cat with gastric chronic fistula. The octapeptide inhibits pentagastrin-induced acid secretion in all three models. Compared to oxyntomodulin, the parent hormone, the synthetic peptide was approximately 150 times less potent but has the same efficacy. Biological data are presented and discussed.

  7. Nuclear nonproliferation and safety: Challenges facing the International Atomic Energy Agency

    SciTech Connect

    Not Available

    1993-09-01

    The Chairman of the Senate Committee on Govermental Affairs asked the United States General Accounting Office (GAO) to review the safeguards and nuclear power plant safety programs of the International Atomic Energy Agency (IAEA). This report examines (1) the effectiveness of IAEA`s safeguards program and the adequacy of program funding, (2) the management of U.S. technical assistance to the IAEA`s safeguards program, and (3) the effectiveness of IAEA`s program for advising United Nations (UN) member states about nuclear power plant safety and the adequacy of program funding. Under its statute and the Treaty on the Non-Proliferation of Nuclear Weapons, IAEA is mandated to administer safeguards to detect diversions of significant quantities of nuclear material from peaceful uses. Because of limits on budget growth and unpaid contributions, IAEA has had difficulty funding the safeguards program. IAEA also conducts inspections of facilities or locations containing declared nuclear material, and manages a program for reviewing the operational safety of designated nuclear power plants. The U.S. technical assistance program for IAEA safeguards, overseen by an interagency coordinating committee, has enhanced the agency`s inspection capabilities, however, some weaknesses still exist. Despite financial limitations, IAEA is meeting its basic safety advisory responsibilities for advising UN member states on nuclear safety and providing requested safety services. However, IAEA`s program for reviewing the operational safety of nuclear power plants has not been fully effective because the program is voluntary and UN member states have not requested IAEA`s review of all nuclear reactors with serious problems. GAO believes that IAEA should have more discretion in selecting reactors for review.

  8. NASA safety program activities in support of the Space Exploration Initiatives Nuclear Propulsion program

    NASA Technical Reports Server (NTRS)

    Sawyer, J. C., Jr.

    1993-01-01

    The activities of the joint NASA/DOE/DOD Nuclear Propulsion Program Technical Panels have been used as the basis for the current development of safety policies and requirements for the Space Exploration Initiatives (SEI) Nuclear Propulsion Technology development program. The Safety Division of the NASA Office of Safety and Mission Quality has initiated efforts to develop policies for the safe use of nuclear propulsion in space through involvement in the joint agency Nuclear Safety Policy Working Group (NSPWG), encouraged expansion of the initial policy development into proposed programmatic requirements, and suggested further expansion into the overall risk assessment and risk management process for the NASA Exploration Program. Similar efforts are underway within the Department of Energy to ensure the safe development and testing of nuclear propulsion systems on Earth. This paper describes the NASA safety policy related to requirements for the design of systems that may operate where Earth re-entry is a possibility. The expected plan of action is to support and oversee activities related to the technology development of nuclear propulsion in space, and support the overall safety and risk management program being developed for the NASA Exploration Program.

  9. NASA safety program activities in support of the Space Exploration Initiatives Nuclear Propulsion program

    NASA Technical Reports Server (NTRS)

    Sawyer, J. C., Jr.

    1993-01-01

    The activities of the joint NASA/DOE/DOD Nuclear Propulsion Program Technical Panels have been used as the basis for the current development of safety policies and requirements for the Space Exploration Initiatives (SEI) Nuclear Propulsion Technology development program. The Safety Division of the NASA Office of Safety and Mission Quality has initiated efforts to develop policies for the safe use of nuclear propulsion in space through involvement in the joint agency Nuclear Safety Policy Working Group (NSPWG), encouraged expansion of the initial policy development into proposed programmatic requirements, and suggested further expansion into the overall risk assessment and risk management process for the NASA Exploration Program. Similar efforts are underway within the Department of Energy to ensure the safe development and testing of nuclear propulsion systems on Earth. This paper describes the NASA safety policy related to requirements for the design of systems that may operate where Earth re-entry is a possibility. The expected plan of action is to support and oversee activities related to the technology development of nuclear propulsion in space, and support the overall safety and risk management program being developed for the NASA Exploration Program.

  10. Integrated Response Time Evaluation Methodology for the Nuclear Safety Instrumentation System

    NASA Astrophysics Data System (ADS)

    Lee, Chang Jae; Yun, Jae Hee

    2017-06-01

    Safety analysis for a nuclear power plant establishes not only an analytical limit (AL) in terms of a measured or calculated variable but also an analytical response time (ART) required to complete protective action after the AL is reached. If the two constraints are met, the safety limit selected to maintain the integrity of physical barriers used for preventing uncontrolled radioactivity release will not be exceeded during anticipated operational occurrences and postulated accidents. Setpoint determination methodologies have actively been developed to ensure that the protective action is initiated before the process conditions reach the AL. However, regarding the ART for a nuclear safety instrumentation system, an integrated evaluation methodology considering the whole design process has not been systematically studied. In order to assure the safety of nuclear power plants, this paper proposes a systematic and integrated response time evaluation methodology that covers safety analyses, system designs, response time analyses, and response time tests. This methodology is applied to safety instrumentation systems for the advanced power reactor 1400 and the optimized power reactor 1000 nuclear power plants in South Korea. The quantitative evaluation results are provided herein. The evaluation results using the proposed methodology demonstrate that the nuclear safety instrumentation systems fully satisfy corresponding requirements of the ART.

  11. Safety and Nonsafety Communications and Interactions in International Nuclear Power Plants

    SciTech Connect

    Kisner, Roger A; Mullens, James Allen; Wilson, Thomas L; Wood, Richard Thomas; Korsah, Kofi; Qualls, A L; Muhlheim, Michael David; Holcomb, David Eugene; Loebl, Andy

    2007-08-01

    Current industry and NRC guidance documents such as IEEE 7-4.3.2, Reg. Guide 1.152, and IEEE 603 do not sufficiently define a level of detail for evaluating interdivisional communications independence. The NRC seeks to establish criteria for safety systems communications that can be uniformly applied in evaluation of a variety of safety system designs. This report focuses strictly on communication issues related to data sent between safety systems and between safety and nonsafety systems. Further, the report does not provide design guidance for communication systems nor present detailed failure modes and effects analysis (FMEA) results for existing designs. This letter report describes communications between safety and nonsafety systems in nuclear power plants outside the United States. A limited study of international nuclear power plants was conducted to ascertain important communication implementations that might have bearing on systems proposed for licensing in the United States. This report provides that following information: 1.communications types and structures used in a representative set of international nuclear power reactors, and 2.communications issues derived from standards and other source documents relevant to safety and nonsafety communications. Topics that are discussed include the following: communication among redundant safety divisions, communications between safety divisions and nonsafety systems, control of safety equipment from a nonsafety workstation, and connection of nonsafety programming, maintenance, and test equipment to redundant safety divisions during operation. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.

  12. Regulatory aspects of nuclear criticality safety in Germany

    SciTech Connect

    Schweer, H.H.

    1996-12-31

    The Atomic Energy Act on the peaceful use of nuclear energy and of the protection against its hazards was Passed in the German parliament in 1959. One of the purposes of this act is {open_quotes}to promote the research, development and utilization of nuclear energy for peaceful purposes.{close_quotes} This act defines fissile nuclear material (Kernbrennstoffe) and lays down the conditions and responsibilities for licensing transportation, storage, and other nuclear facilities including reactors. Based on the Atomic Energy Act, the ordinance for radiation protection was passed in October 1976. This ordinance contains requirements concerning the handling and transport of radioactive materials and basic principles for radiation protection.

  13. 75 FR 53985 - Southern Nuclear Operating Company Establishment of Atomic Safety And Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-02

    ... a Sustainable Coast, and Georgia Women's Action for New Directions for Clean Energy.\\1\\ \\1\\ On May... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Southern Nuclear Operating Company Establishment of Atomic Safety And Licensing Board Pursuant to...

  14. ASME Nuclear Crane Standards for Enhanced Crane Safety and Increased Profit

    NASA Astrophysics Data System (ADS)

    Parkhurst, Stephen N.

    2000-01-01

    The ASME NOG-1 standard, 'Rules for Construction of Overhead and Gantry Cranes', covers top running cranes for nuclear facilities; with the ASME NUM-1 standard, 'Rules for Construction of Cranes, Monorails, and Hoists', covering the single girder, underhung, wall and jib cranes, as well as the monorails and hoists. These two ASME nuclear crane standards provide criteria for designing, inspecting and testing overhead handling equipment with enhanced safety to meet the 'defense-in-depth' approach of the United States Nuclear Regulatory Commission (USNRC) documents NUREG 0554 and NUREG 0612. In addition to providing designs for enhanced safety, the ASME nuclear crane standards provide a basis for purchasing overhead handling equipment with standard safety features, based upon accepted engineering principles, and including performance and environmental parameters specific to nuclear facilities. The ASME NOG-1 and ASME NUM-1 standards not only provide enhanced safety for handling a critical load, but also increase profit by minimizing the possibility of load drops, by reducing cumbersome operating restrictions, and by providing the foundation for a sound licensing position. The ASME nuclear crane standards can also increase profit by providing the designs and information to help ensure that the right standard equipment is purchased. Additionally, the ASME nuclear crane standards can increase profit by providing designs and information to help address current issues, such as the qualification of nuclear plant cranes for making 'planned engineered lifts' for steam generator replacement and decommissioning.

  15. Management concepts and safety applications for nuclear fuel facilities

    SciTech Connect

    Eisner, H.; Scotti, R.S.; Delicate, W.S.

    1995-05-01

    This report presents an overview of effectiveness of management control of safety. It reviews several modern management control theories as well as the general functions of management and relates them to safety issues at the corporate and at the process safety management (PSM) program level. Following these discussions, structured technique for assessing management of the safety function is suggested. Seven modern management control theories are summarized, including business process reengineering, the learning organization, capability maturity, total quality management, quality assurance and control, reliability centered maintenance, and industrial process safety. Each of these theories is examined for-its principal characteristics and implications for safety management. The five general management functions of planning, organizing, directing, monitoring, and integrating, which together provide control over all company operations, are discussed. Under the broad categories of Safety Culture, Leadership and Commitment, and Operating Excellence, key corporate safety elements and their subelements are examined. The three categories under which PSM program-level safety issues are described are Technology, Personnel, and Facilities.

  16. Potential safety-related incidents with possible applicability to a nuclear fuel reprocessing plant

    SciTech Connect

    Durant, W.S.; Perkins, W.C.; Lee, R.; Stoddard, D.H.

    1982-05-20

    The Safety Technology Group is developing methodology that can be used to assess the risk of operating a plant to reprocess spent nuclear fuel. As an early step in the methodology, a preliminary hazards analysis identifies safety-related incidents. In the absence of appropriate safety features, these incidents could lead to significant consequences and risk to onsite personnel or to the public. This report is a compilation of potential safety-related incidents that have been identified in studies at SRL and in safety analyses of various commercially designed reprocessing plants. It is an expanded revision of the version originally published as DP-1558, Published December 1980.

  17. IMPLEMENTATION OF DEFENSE NUCLEAR FACILITY SAFETY BOARD RECOMMENDATION 2000-2 AT WIPP

    SciTech Connect

    Jackson, K.; Wu, C.

    2002-02-26

    The Defense Nuclear Safeties Board (DNFSB) issued Recommendation 2000-2 on March 8, 2000, concerning the degrading conditions of vital safety systems, or systems important to nuclear safety, at DOE sites across the nation. The Board recommended that the DOE take action to assess the condition of its nuclear systems to ensure continued operational readiness of vital safety systems that are important for safely accomplishing the DOE's mission. To verify the readiness of vital safety systems, a two-phased approach was established. Phase I consisted of a qualitative assessment to approved criteria of the defined vital safety systems by operating contractor personnel, overseen by Federal field office personnel. Based on Phase I Assessment results, vital safety systems with significant deficiencies would be further assessed in Phase II, a more extensive quantitative assessment, by a contractor and Federal team, using a second set of criteria. In addition, Defense Nuclear Facility Safety Board Recommendation 2000-2 concluded that the degradation of confinement ventilation systems was of major concern, and issued a separate set of criteria to perform a Phase II Assessment on confinement ventilation systems.

  18. 78 FR 47014 - Configuration Management Plans for Digital Computer Software Used in Safety Systems of Nuclear...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-02

    ... COMMISSION Configuration Management Plans for Digital Computer Software Used in Safety Systems of Nuclear... 1 of RG 1.169, ``Configuration Management Plans for Digital Computer Software Used in Safety Systems... those systems include software. This RG is one of six RG revisions addressing computer software...

  19. Current state of nuclear fuel cycles in nuclear engineering and trends in their development according to the environmental safety requirements

    NASA Astrophysics Data System (ADS)

    Vislov, I. S.; Pischulin, V. P.; Kladiev, S. N.; Slobodyan, S. M.

    2016-08-01

    The state and trends in the development of nuclear fuel cycles in nuclear engineering, taking into account the ecological aspects of using nuclear power plants, are considered. An analysis of advantages and disadvantages of nuclear engineering, compared with thermal engineering based on organic fuel types, was carried out. Spent nuclear fuel (SNF) reprocessing is an important task in the nuclear industry, since fuel unloaded from modern reactors of any type contains a large amount of radioactive elements that are harmful to the environment. On the other hand, the newly generated isotopes of uranium and plutonium should be reused to fabricate new nuclear fuel. The spent nuclear fuel also includes other types of fission products. Conditions for SNF handling are determined by ecological and economic factors. When choosing a certain handling method, one should assess these factors at all stages of its implementation. There are two main methods of SNF handling: open nuclear fuel cycle, with spent nuclear fuel assemblies (NFAs) that are held in storage facilities with their consequent disposal, and closed nuclear fuel cycle, with separation of uranium and plutonium, their purification from fission products, and use for producing new fuel batches. The development of effective closed fuel cycles using mixed uranium-plutonium fuel can provide a successful development of the nuclear industry only under the conditions of implementation of novel effective technological treatment processes that meet strict requirements of environmental safety and reliability of process equipment being applied. The diversity of technological processes is determined by different types of NFA devices and construction materials being used, as well as by the composition that depends on nuclear fuel components and operational conditions for assemblies in the nuclear power reactor. This work provides an overview of technological processes of SNF treatment and methods of handling of nuclear fuel

  20. 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.

  1. Implementation plan for the Defense Nuclear Facilities Safety Board Recommendation 90-7. Revision 1

    SciTech Connect

    Borsheim, G.L.; Cash, R.J.; Dukelow, G.T.

    1992-12-01

    This document revises the original plan submitted in March 1991 for implementing the recommendations made by the Defense Nuclear Facilities Safety Board in their Recommendation 90-7 to the US Department of Energy. Recommendation 90-7 addresses safety issues of concern for 24 single-shell, high-level radioactive waste tanks containing ferrocyanide compounds at the Hanford Site. The waste in these tanks is a potential safety concern because, under certain conditions involving elevated temperatures and low concentrations of nonparticipating diluents, ferrocyanide compounds in the presence of oxidizing materials can undergo a runaway (propagating) chemical reaction. This document describes those activities underway by the Hanford Site contractor responsible for waste tank safety that address each of the six parts of Defense Nuclear Facilities Safety Board Recommendation 90-7. This document also identifies the progress made on these activities since the beginning of the ferrocyanide safety program in September 1990. Revised schedules for planned activities are also included.

  2. 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.

  3. Improving the regulation of safety at DOE nuclear facilities. Final report: Appendices

    SciTech Connect

    1995-12-01

    The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

  4. Improving the regulation of safety at DOE nuclear facilities. Final report

    SciTech Connect

    1995-12-01

    The report strongly recommends that, with the end of the Cold War, safety and health at DOE facilities should be regulated by outside agencies rather than by DOE itself. The three major recommendations are: under any regulatory scheme, DOE must maintain a strong internal safety management system; essentially all aspects of safety at DOE`s nuclear facilities should be externally regulated; and existing agencies rather than a new one should be responsible for external regulation.

  5. 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)

  6. Safety analysis of the nuclear chemistry Building 151

    SciTech Connect

    Kvam, D.

    1984-06-29

    This report summarizes the results of a safety analysis that was done on Building 151. The report outlines the methodology, the analysis, and the findings that led to the low hazard classification. No further safety evaluation is indicated at this time. 5 tables.

  7. 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)

  8. WASTE PROCESSING ANNUAL NUCLEAR SAFETY RELATED R AND D REPORT FOR CY2008

    SciTech Connect

    Fellinger, A.

    2009-10-15

    The Engineering and Technology Office of Waste Processing identifies and reduces engineering and technical risks associated with key waste processing project decisions. The risks, and actions taken to mitigate those risks, are determined through technology readiness assessments, program reviews, technology information exchanges, external technical reviews, technical assistance, and targeted technology development and deployment (TDD). The Office of Waste Processing TDD program prioritizes and approves research and development scopes of work that address nuclear safety related to processing of highly radioactive nuclear wastes. Thirteen of the thirty-five R&D approved work scopes in FY2009 relate directly to nuclear safety, and are presented in this report.

  9. Nuclear criticality safety evaluation of large cylinder cleaning operations in X-705, Portsmouth Gaseous diffusion Plant

    SciTech Connect

    Sheaffer, M.K.; Keeton, S.C.; Lutz, H.F.

    1995-06-01

    This report evaluates nuclear criticality safety for large cylinder cleaning operations in the Decontamination and Recovery Facility, X-705, at the Portsmouth Gaseous Diffusion Plant. A general description of current cleaning procedures and required hardware/equipment is presented, and documentation for large cylinder cleaning operations is identified and described. Control parameters, design features, administrative controls, and safety systems relevant to nuclear criticality are discussed individually, followed by an overall assessment based on the Double Contingency Principle. Recommendations for enhanced safety are suggested, and issues for increased efficiency are presented.

  10. First observation of Hb D-Ouled Rabah [beta19(B1)Asn>Lys] in the Turkish population.

    PubMed

    Köseler, Aylin; Bahadır, Anzel; Koyuncu, Hasan; Atalay, Ayfer; Atalay, Erol Ömer

    2008-03-05

    Hb D-Ouled Rabah [beta19(B1)Asn>Lys] is a rare hemoglobin (Hb) beta chain variant reported from Tuareg tribes in Algeria and once from China. It was suggested that Hb D-Ouled Rabah might be specific of Berber-speaking populations. Our report describes the first observation of this hemoglobin variant in the Turkish population.

  11. Characterization and improvement of the nuclear safety culture through self-assessment

    SciTech Connect

    Levin, H.A.; McGehee, R.B.; Cottle, W.T.

    1996-12-31

    Organizational culture has a powerful influence on overall corporate performance. The ability to sustain superior results in ensuring the public`s health and safety is predicated on an organization`s deeply embedded values and behavioral norms and how these affect the ability to change and seek continuous improvement. The nuclear industry is developing increased recognition of the relationship of culture to nuclear safety performance as a critical element of corporate strategy. This paper describes a self-assessment methodology designed to characterize and improve the nuclear safety culture, including processes for addressing employee concerns. This methodology has been successfully applied on more than 30 occasions in the last several years, resulting in measurable improvements in safety performance and quality and employee motivation, productivity, and morale. Benefits and lessons learned are also presented.

  12. Uncertainty Quantification for Safety Verification Applications in Nuclear Power Plants

    NASA Astrophysics Data System (ADS)

    Boafo, Emmanuel

    There is an increasing interest in computational reactor safety analysis to systematically replace the conservative calculations by best estimate calculations augmented by quantitative uncertainty analysis methods. This has been necessitated by recent regulatory requirements that have permitted the use of such methods in reactor safety analysis. Stochastic uncertainty quantification methods have shown great promise, as they are better suited to capture the complexities in real engineering problems. This study proposes a framework for performing uncertainty quantification based on the stochastic approach, which can be applied to enhance safety analysis. (Abstract shortened by ProQuest.).

  13. 77 FR 50722 - Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-22

    ... COMMISSION Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants... entitled ``Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power... systems of nuclear power plants. In particular, the methods are consistent with part 50 of Title 10 of the...

  14. Radiation safety audit of a high volume Nuclear Medicine Department.

    PubMed

    Jha, Ashish Kumar; Singh, Abhijith Mohan; Shetye, Bhakti; Shah, Sneha; Agrawal, Archi; Purandare, Nilendu Chandrakant; Monteiro, Priya; Rangarajan, Venkatesh

    2014-10-01

    Professional radiation exposure cannot be avoided in nuclear medicine practices. It can only be minimized up to some extent by implementing good work practices. The aim of our study was to audit the professional radiation exposure and exposure rate of radiation worker working in and around Department of nuclear medicine and molecular imaging, Tata Memorial Hospital. We calculated the total number of nuclear medicine and positron emission tomography/computed tomography (PET/CT) procedures performed in our department and the radiation exposure to the radiation professionals from year 2009 to 2012. We performed an average of 6478 PET/CT scans and 3856 nuclear medicine scans/year from January 2009 to December 2012. The average annual whole body radiation exposure to nuclear medicine physician, technologist and nursing staff are 1.74 mSv, 2.93 mSv and 4.03 mSv respectively. Efficient management and deployment of personnel is of utmost importance to optimize radiation exposure in a high volume nuclear medicine setup in order to work without anxiety of high radiation exposure.

  15. Radiation safety audit of a high volume Nuclear Medicine Department

    PubMed Central

    Jha, Ashish Kumar; Singh, Abhijith Mohan; Shetye, Bhakti; Shah, Sneha; Agrawal, Archi; Purandare, Nilendu Chandrakant; Monteiro, Priya; Rangarajan, Venkatesh

    2014-01-01

    Introduction: Professional radiation exposure cannot be avoided in nuclear medicine practices. It can only be minimized up to some extent by implementing good work practices. Aim and Objectives: The aim of our study was to audit the professional radiation exposure and exposure rate of radiation worker working in and around Department of nuclear medicine and molecular imaging, Tata Memorial Hospital. Materials and Methods: We calculated the total number of nuclear medicine and positron emission tomography/computed tomography (PET/CT) procedures performed in our department and the radiation exposure to the radiation professionals from year 2009 to 2012. Results: We performed an average of 6478 PET/CT scans and 3856 nuclear medicine scans/year from January 2009 to December 2012. The average annual whole body radiation exposure to nuclear medicine physician, technologist and nursing staff are 1.74 mSv, 2.93 mSv and 4.03 mSv respectively. Conclusion: Efficient management and deployment of personnel is of utmost importance to optimize radiation exposure in a high volume nuclear medicine setup in order to work without anxiety of high radiation exposure. PMID:25400361

  16. Digital Signal Processing Methods for Safety Systems Employed in Nuclear Power Industry

    NASA Astrophysics Data System (ADS)

    Popescu, George

    Some of the major safety concerns in the nuclear power industry focus on the readiness of nuclear power plant safety systems to respond to an abnormal event, the security of special nuclear materials in used nuclear fuels, and the need for physical security to protect personnel and reactor safety systems from an act of terror. Routine maintenance and tests of all nuclear reactor safety systems are performed on a regular basis to confirm the ability of these systems to operate as expected. However, these tests do not determine the reliability of these safety systems and whether the systems will perform for the duration of an accident and whether they will perform their tasks without failure after being engaged. This research has investigated the progression of spindle asynchronous error motion determined from spindle accelerations to predict bearings failure onset. This method could be applied to coolant pumps that are essential components of emergency core cooling systems at all nuclear power plants. Recent security upgrades mandated by the Nuclear Regulatory Commission and the Department of Homeland Security have resulted in implementation of multiple physical security barriers around all of the commercial and research nuclear reactors in the United States. A second part of this research attempts to address an increased concern about illegal trafficking of Special Nuclear Materials (SNM). This research describes a multi element scintillation detector system designed for non - invasive (passive) gamma ray surveillance for concealed SNM that may be within an area or sealed in a package, vehicle or shipping container. Detection capabilities of the system were greatly enhanced through digital signal processing, which allows the combination of two very powerful techniques: 1) Compton Suppression (CS) and 2) Pulse Shape Discrimination (PSD) with less reliance on complicated analog instrumentation.

  17. Evaluating software for safety systems in nuclear power plants

    SciTech Connect

    Lawrence, J.D.; Persons, W.L.; Preckshot, G.G.; Gallagher, J.

    1994-01-11

    In 1991, LLNL was asked by the NRC to provide technical assistance in various aspects of computer technology that apply to computer-based reactor protection systems. This has involved the review of safety aspects of new reactor designs and the provision of technical advice on the use of computer technology in systems important to reactor safety. The latter includes determining and documenting state-of-the-art subjects that require regulatory involvement by the NRC because of their importance in the development and implementation of digital computer safety systems. These subjects include data communications, formal methods, testing, software hazards analysis, verification and validation, computer security, performance, software complexity and others. One topic software reliability and safety is the subject of this paper.

  18. Framework for Integrating Safety, Operations, Security, and Safeguards in the Design and Operation of Nuclear Facilities

    SciTech Connect

    Darby, John L.; Horak, Karl Emanuel; LaChance, Jeffrey L.; Tolk, Keith Michael; Whitehead, Donnie Wayne

    2007-10-01

    The US is currently on the brink of a nuclear renaissance that will result in near-term construction of new nuclear power plants. In addition, the Department of Energy’s (DOE) ambitious new Global Nuclear Energy Partnership (GNEP) program includes facilities for reprocessing spent nuclear fuel and reactors for transmuting safeguards material. The use of nuclear power and material has inherent safety, security, and safeguards (SSS) concerns that can impact the operation of the facilities. Recent concern over terrorist attacks and nuclear proliferation led to an increased emphasis on security and safeguard issues as well as the more traditional safety emphasis. To meet both domestic and international requirements, nuclear facilities include specific SSS measures that are identified and evaluated through the use of detailed analysis techniques. In the past, these individual assessments have not been integrated, which led to inefficient and costly design and operational requirements. This report provides a framework for a new paradigm where safety, operations, security, and safeguards (SOSS) are integrated into the design and operation of a new facility to decrease cost and increase effectiveness. Although the focus of this framework is on new nuclear facilities, most of the concepts could be applied to any new, high-risk facility.

  19. Survey of systems safety analysis methods and their application to nuclear waste management systems

    SciTech Connect

    Pelto, P.J.; Winegardner, W.K.; Gallucci, R.H.V.

    1981-11-01

    This report reviews system safety analysis methods and examines their application to nuclear waste management systems. The safety analysis methods examined include expert opinion, maximum credible accident approach, design basis accidents approach, hazard indices, preliminary hazards analysis, failure modes and effects analysis, fault trees, event trees, cause-consequence diagrams, G0 methodology, Markov modeling, and a general category of consequence analysis models. Previous and ongoing studies on the safety of waste management systems are discussed along with their limitations and potential improvements. The major safety methods and waste management safety related studies are surveyed. This survey provides information on what safety methods are available, what waste management safety areas have been analyzed, and what are potential areas for future study.

  20. Environmental safety aspects of the new spent nuclear fuel management and storage system at Ignalina NPP

    SciTech Connect

    Poskas, P.; Ragaisis, V.; Adomaitis, J. E.

    2007-07-01

    In the framework of the preparation for the decommissioning of the Ignalina Nuclear Power Plant (INPP) a new Interim Spent Nuclear Fuel Storage Facility (ISFSF) will be built in the existing sanitary protection zone (SPZ) of INPP. In addition to the ISFSF, the new spent nuclear fuel management activity will include all necessary spent nuclear fuel retrieval and packaging operations at the Reactor Units, transfer of storage casks to the ISFSF, and other activities appropriate to the chosen design solution and required for the safe removal of the existing spent nuclear fuel from storage pools and insertion into the new ISFSF. The Republic of Lithuania regulations require that the average annual dose to the critical group members of population due to operation of nuclear facility shall not exceed dose constraint. If several nuclear facilities are located in the same SPZ, the same dose constraint shall envelope radiological impacts from all operating and planned nuclear facilities. The paper discusses radiological safety assessment aspects as relevant for the new nuclear activity to be implemented in the SPZ of INPP considering specificity of Lithuanian regulatory requirements. The safety assessment methodology aspects, results and conclusions as concern public exposure are outlined and discussed. (authors)

  1. Nuclear criticality safety staff training and qualifications at Los Alamos National Laboratory

    SciTech Connect

    Monahan, S.P.; McLaughlin, T.P.

    1997-05-01

    Operations involving significant quantities of fissile material have been conducted at Los Alamos National Laboratory continuously since 1943. Until the advent of the Laboratory`s Nuclear Criticality Safety Committee (NCSC) in 1957, line management had sole responsibility for controlling criticality risks. From 1957 until 1961, the NCSC was the Laboratory body which promulgated policy guidance as well as some technical guidance for specific operations. In 1961 the Laboratory created the position of Nuclear Criticality Safety Office (in addition to the NCSC). In 1980, Laboratory management moved the Criticality Safety Officer (and one other LACEF staff member who, by that time, was also working nearly full-time on criticality safety issues) into the Health Division office. Later that same year the Criticality Safety Group, H-6 (at that time) was created within H-Division, and staffed by these two individuals. The training and education of these individuals in the art of criticality safety was almost entirely self-regulated, depending heavily on technical interactions between each other, as well as NCSC, LACEF, operations, other facility, and broader criticality safety community personnel. Although the Los Alamos criticality safety group has grown both in size and formality of operations since 1980, the basic philosophy that a criticality specialist must be developed through mentoring and self motivation remains the same. Formally, this philosophy has been captured in an internal policy, document ``Conduct of Business in the Nuclear Criticality Safety Group.`` There are no short cuts or substitutes in the development of a criticality safety specialist. A person must have a self-motivated personality, excellent communications skills, a thorough understanding of the principals of neutron physics, a safety-conscious and helpful attitude, a good perspective of real risk, as well as a detailed understanding of process operations and credible upsets.

  2. Results of operation and current safety performance of nuclear facilities located in the Russian Federation

    NASA Astrophysics Data System (ADS)

    Kuznetsov, V. M.; Khvostova, M. S.

    2016-12-01

    After the NPP radiation accidents in Russia and Japan, a safety statu of Russian nuclear power plants causes concern. A repeated life time extension of power unit reactor plants, designed at the dawn of the nuclear power engineering in the Soviet Union, power augmentation of the plants to 104-109%, operation of power units in a daily power mode in the range of 100-70-100%, the use of untypical for NPP remixed nuclear fuel without a careful study of the results of its application (at least after two operating periods of the research nuclear installations), the aging of operating personnel, and many other management actions of the State Corporation "Rosatom", should attract the attention of the Federal Service for Ecological, Technical and Atomic Supervision (RosTekhNadzor), but this doesn't happen. The paper considers safety issues of nuclear power plants operating in the Russian Federation. The authors collected statistical information on violations in NPP operation over the past 25 years, which shows that even after repeated relaxation over this period of time of safety regulation requirements in nuclear industry and highly expensive NPP modernization, the latter have not become more safe, and the statistics confirms this. At a lower utilization factor high-power pressure-tube reactors RBMK-1000, compared to light water reactors VVER-440 and 1000, have a greater number of violations and that after annual overhauls. A number of direct and root causes of NPP mulfunctions is still high and remains stable for decades. The paper reveals bottlenecks in ensuring nuclear and radiation safety of nuclear facilities. Main outstanding issues on the storage of spent nuclear fuel are defined. Information on emissions and discharges of radioactive substances, as well as fullness of storages of solid and liquid radioactive waste, located at the NPP sites are presented. Russian NPPs stress test results are submitted, as well as data on the coming removal from operation of NPP

  3. Safety Software Guide Perspectives for the Design of New Nuclear Facilities (U)

    SciTech Connect

    VINCENT, Andrew

    2005-07-14

    In June of this year, the Department of Energy (DOE) issued directives DOE O 414.1C and DOE G 414.1-4 to improve quality assurance programs, processes, and procedures among its safety contractors. Specifically, guidance entitled, ''Safety Software Guide for use with 10 CFR 830 Subpart A, Quality Assurance Requirements, and DOE O 414.1C, Quality Assurance, DOE G 414.1-4'', provides information and acceptable methods to comply with safety software quality assurance (SQA) requirements. The guidance provides a roadmap for meeting DOE O 414.1C, ''Quality Assurance'', and the quality assurance program (QAP) requirements of Title 10 Code of Federal Regulations (CFR) 830, Subpart A, Quality Assurance, for DOE nuclear facilities and software application activities. [1, 2] The order and guide are part of a comprehensive implementation plan that addresses issues and concerns documented in Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2002-1. [3] Safety SQA requirements for DOE as well as National Nuclear Security Administration contractors are necessary to implement effective quality assurance (QA) processes and achieve safe nuclear facility operations. DOE G 414.1-4 was developed to provide guidance on establishing and implementing effective QA processes tied specifically to nuclear facility safety software applications. The Guide includes software application practices covered by appropriate national and international consensus standards and various processes currently in use at DOE facilities. While the safety software guidance is considered to be of sufficient rigor and depth to ensure acceptable reliability of safety software at all DOE nuclear facilities, new nuclear facilities are well suited to take advantage of the guide to ensure compliant programs and processes are implemented. Attributes such as the facility life-cycle stage and the hazardous nature of each facility operations are considered, along with the category and level of importance of the

  4. Preliminary nuclear safety assessment of the NEPST (Topaz 2) space reactor program

    NASA Astrophysics Data System (ADS)

    Marshall, A. C.

    The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz 2 space nuclear power system. A preliminary nuclear safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz 2 activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclear safety assessment included a number of deterministic analyses, such as the following: neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz 2 system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

  5. Preliminary nuclear safety assessment of the NEPST (Topaz II) space reactor program

    SciTech Connect

    Marshall, A.C.

    1993-01-01

    The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz II space nuclear power system. A preliminary nuclear safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclear safety assessment included a number of deterministic analyses, such as; neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz II system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

  6. Space nuclear safety program: Progress report, July--September 1987

    SciTech Connect

    George, T.G.

    1989-02-01

    This quarterly report describes studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems, carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. The studies discussed are ongoing; the results and conclusions described may change as the work progresses. 20 figs., 4 tabs.

  7. Space nuclear safety program. Progress report, January 1984

    SciTech Connect

    Bronisz, S.E.

    1984-07-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

  8. Space Nuclear Safety Program: Progress report, January-March 1987

    SciTech Connect

    Lewin, R.; George, T.G.

    1988-07-01

    This quarterly report describes studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems, which were carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work progresses.

  9. Space Nuclear-Safety Program progress report, February 1983

    SciTech Connect

    Bronisz, S.E.

    1983-08-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions may change as the work continues.

  10. Space nuclear-safety program. Progress report, October 1982

    SciTech Connect

    Bronisz, S.E.

    1983-03-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

  11. Space nuclear safety program. Progress report, December 1982

    SciTech Connect

    Bronisz, S.E.

    1983-06-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

  12. Space nuclear safety program. Progress report, October 1983

    SciTech Connect

    Bronisz, S.E.

    1984-03-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory.

  13. Space nuclear safety program. Progress report, August 1983

    SciTech Connect

    Bronisz, S.E.

    1984-01-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

  14. Space nuclear-safety program, November 1982. Progress report

    SciTech Connect

    Bronisz, S.E.

    1983-05-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

  15. Space Nuclear Safety Program. Progress report, June 1984

    SciTech Connect

    George, T.G.

    1985-11-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work continues. 36 figs.

  16. Space Nuclear Safety Program. Progress report, August 1984

    SciTech Connect

    George, T.G.

    1985-11-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work progresses. 41 figs.

  17. Space nuclear safety program. Progress report, October-December 1984

    SciTech Connect

    George, T.G.

    1986-05-01

    This quarterly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed are ongoing; the results and conclusions described may change as the work progresses.

  18. Space nuclear safety program. Progress report, July 1983

    SciTech Connect

    Bronisz, S.E.

    1983-11-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Most of the studies discussed here are ongoing. Results and conclusions described may change as the work continues.

  19. Space Nuclear Safety Program. Progress report, May 1984

    SciTech Connect

    George, T.G.

    1985-09-01

    This technical monthly report covers studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems carried out for the Office of Special Nuclear Projects of the US Department of Energy by Los Alamos National Laboratory. Covered are: general-purpose heat source testing, light-weight radioisotope heater unit, and iridium biaxial testing.

  20. 10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... important to safety for nuclear power plants. 50.49 Section 50.49 Energy NUCLEAR REGULATORY COMMISSION... nuclear power plants. (a) Each holder of or an applicant for an operating license issued under this part... nuclear power plant for which the certifications required under § 50.82(a)(1) or § 52.110(a)(1) of...

  1. 10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... important to safety for nuclear power plants. 50.49 Section 50.49 Energy NUCLEAR REGULATORY COMMISSION... nuclear power plants. (a) Each holder of or an applicant for an operating license issued under this part... nuclear power plant for which the certifications required under § 50.82(a)(1) or § 52.110(a)(1) of this...

  2. 10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... important to safety for nuclear power plants. 50.49 Section 50.49 Energy NUCLEAR REGULATORY COMMISSION... nuclear power plants. (a) Each holder of or an applicant for an operating license issued under this part... nuclear power plant for which the certifications required under § 50.82(a)(1) or § 52.110(a)(1) of this...

  3. 10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... important to safety for nuclear power plants. 50.49 Section 50.49 Energy NUCLEAR REGULATORY COMMISSION... nuclear power plants. (a) Each holder of or an applicant for an operating license issued under this part... nuclear power plant for which the certifications required under § 50.82(a)(1) or § 52.110(a)(1) of this...

  4. 10 CFR 50.49 - Environmental qualification of electric equipment important to safety for nuclear power plants.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... important to safety for nuclear power plants. 50.49 Section 50.49 Energy NUCLEAR REGULATORY COMMISSION... nuclear power plants. (a) Each holder of or an applicant for an operating license issued under this part... nuclear power plant for which the certifications required under § 50.82(a)(1) or § 52.110(a)(1) of this...

  5. Space nuclear safety program, May 1983. Progress report

    SciTech Connect

    Bronisz, S.E.

    1983-10-01

    The studies related to the use of /sup 238/PuO/sub 2/ in radioisotope power systems, pertained to the General-Purpose Heat Source (compatibility and safety verification) and to the Light-Weight Radioisotope Heater units (overpressure and impact tests).

  6. Lessons in Nuclear Safety, Panel on Integration of People and Programs

    SciTech Connect

    Pinkston, David

    2015-02-24

    Four slides present a historical perspective on the evolution of nuclear safety, a description of systemic misalignment (available resources do not match expectations, demographic cliff developing, promulgation of increased expectations and new requirements proceeds unabated), and needs facing nuclear safety (financial stability, operational stability, and succession planning). The following conclusions are stated under the heading "Nuclear Safety - 'The System'": the current universe of requirements is too large for the resource pool available; the current universe of requirements has too many different sources of interpretation; there are so many indicators that it’s hard to know what is leading (or important); and the net result can come to defy integrated comprehension at the worker level.

  7. Assessment of radiation safety awareness among nuclear medicine nurses: a pilot study

    NASA Astrophysics Data System (ADS)

    Yunus, N. A.; Abdullah, M. H. R. O.; Said, M. A.; Ch'ng, P. E.

    2014-11-01

    All nuclear medicine nurses need to have some knowledge and awareness on radiation safety. At present, there is no study to address this issue in Malaysia. The aims of this study were (1) to determine the level of knowledge and awareness on radiation safety among nuclear medicine nurses at Putrajaya Hospital in Malaysia and (2) to assess the effectiveness of a training program provided by the hospital to increase the knowledge and awareness of the nuclear medicine nurses. A total of 27 respondents attending a training program on radiation safety were asked to complete a questionnaire. The questionnaire consists 16 items and were categorized into two main areas, namely general radiation knowledge and radiation safety. Survey data were collected before and after the training and were analyzed using descriptive statistics and paired sample t-test. Respondents were scored out of a total of 16 marks with 8 marks for each area. The findings showed that the range of total scores obtained by the nuclear medicine nurses before and after the training were 6-14 (with a mean score of 11.19) and 13-16 marks (with a mean score of 14.85), respectively. Findings also revealed that the mean score for the area of general radiation knowledge (7.59) was higher than that of the radiation safety (7.26). Currently, the knowledge and awareness on radiation safety among the nuclear medicine nurses are at the moderate level. It is recommended that a national study be conducted to assess and increase the level of knowledge and awareness among all nuclear medicine nurses in Malaysia.

  8. Nuclear criticality safety experiments, calculations, and analyses: 1958 to 1982. Volume 1. Lookup tables

    SciTech Connect

    Koponen, B.L.; Hampel, V.E.

    1982-10-21

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

  9. Strengthening safety compliance in nuclear power operations: a role-based approach.

    PubMed

    Martínez-Córcoles, Mario; Gracia, Francisco J; Tomás, Inés; Peiró, José M

    2014-07-01

    Safety compliance is of paramount importance in guaranteeing the safe running of nuclear power plants. However, it depends mostly on procedures that do not always involve the safest outcomes. This article introduces an empirical model based on the organizational role theory to analyze the influence of legitimate sources of expectations (procedures formalization and leadership) on workers' compliance behaviors. The sample was composed of 495 employees from two Spanish nuclear power plants. Structural equation analysis showed that, in spite of some problematic effects of proceduralization (such as role conflict and role ambiguity), procedure formalization along with an empowering leadership style lead to safety compliance by clarifying a worker's role in safety. Implications of these findings for safety research are outlined, as well as their practical implications. © 2014 Society for Risk Analysis.

  10. Safety, Security, and Stability: The Role of Nuclear Control Regimes in a Proliferated World

    DTIC Science & Technology

    1995-05-01

    Harper Collins College, 1993), 107. 6Lewis A. Dunn and Herman Kahn, Trends in Nuclear Proliferation: 1975-1995, Projections, Problems, and Policy...America’s Foreign Policy in a Changing World (New York: Harper Collins College Publishers, 1994), 288. 47 Richard Nixon, The Memoirs of Richard Nixon...Nuclear Weapons Safety, 101st Cong, 2nd Session, No. 15 (Washington: GPO,1990), 9. 136 Ibid, 10. 137 Shaun Gregory, The Hidden Cost of Deterrence

  11. The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine.

    SciTech Connect

    Kot, C.

    1999-05-10

    Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Department of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments.

  12. Safety issues in robotic handling of nuclear weapon parts

    SciTech Connect

    Drotning, W.; Wapman, W.; Fahrenholtz, J.

    1993-12-31

    Robotic systems are being developed by the Intelligent Systems and Robotics Center at Sandia National Laboratories to perform automated handling tasks with radioactive weapon parts. These systems will reduce the occupational radiation exposure to workers by automating operations that are currently performed manually. The robotic systems at Sandia incorporate several levels of mechanical, electrical, and software safety for handling hazardous materials. For example, tooling used by the robot to handle radioactive parts has been designed with mechanical features that allow the robot to release its payload only at designated locations in the robotic workspace. In addition, software processes check for expected and unexpected situations throughout the operations. Incorporation of features such as these provides multiple levels of safety for handling hazardous or valuable payloads with automated intelligent systems.

  13. 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.

  14. Manned space flight nuclear system safety. Volume 7: Literature review. Part 1: Literature search and evaluation

    NASA Technical Reports Server (NTRS)

    1972-01-01

    A review of the literature used in conducting the manned space flight nuclear system safety study is presented. The objectives of the presentation are to identify and evaluate for potential application to study the existing related literature and to provide the information required to include the related literature in the NASA Aerospace Safety Research and Data Institute. More than 15,000 documents were evaluated and identification forms were prepared for 850 reports.

  15. Organizational analysis and safety for utilities with nuclear power plants: perspectives for organizational assessment. Volume 2. [PWR; BWR

    SciTech Connect

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Nadel, M.V.; Scott, W.G.; Connor, P.E.; Kerwin, N.; Kennedy, J.K. Jr.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. Volume 1 of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety. The six chapters of this volume discuss the major elements in our general approach to safety in the nuclear industry. The chapters include information on organizational design and safety; organizational governance; utility environment and safety related outcomes; assessments by selected federal agencies; review of data sources in the nuclear power industry; and existing safety indicators.

  16. Nuclear Health and Safety: Efforts to Strengthen DOE’s Health and Epidemiology Programs

    DTIC Science & Technology

    1991-02-01

    persons poten- tially exposed to radiation, and (4) persons exposed to potential hazards resulting from non- nuclear energy production. The health...Workers Could Be Better Protected at Ohio Defense Plants (QAO/RCED-86.61, Dec. 13, 1985) and Nuclear Health and Safty . Summary o Major Problems at DOE’s...of Energy (March 1990). Page 13 GAO/RCED-91-67 Nuclear Health and Safety Appendix I Reviews of DOE’s Health Programs DOE’S epidemiology activities

  17. Evaluating the safety of aging nuclear reactor pressure vessels

    SciTech Connect

    Pennell, W.E.

    1996-05-01

    Regulatory requirements limit the permissible accumulation of irradiation damage in RPV material such that adequate fracture prevention margins are maintained throughout the licensed operating period of a nuclear plant. Experience with application of those requirements has identified a number of areas where they could be further refined to eliminate excess conservatism. Research is ongoin to provide the data required to support refinement of the regulatory requirements. Research programs are investigating theeffects of local brittle zones, shallow flaws, biaxial loading, and stainless steel cladding. Preliminary results from this research indicate a potential for beneficial changes in the P-T curve and PTS analysis rules.

  18. 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.

  19. Nuclear reactor safety. Progress report, October 1-December 31, 1980

    SciTech Connect

    Stevenson, M.G.; Vigil, J.C.

    1981-09-01

    Development of the fast-running Transient Reactor Analysis Code (TRAC) version (PF1) continued during the quarter with numerical improvements and addition of a stratified-flow model. Independent assessment of the detailed version (PD2) continued with several Loss-Of-Fluid Test (LOFT) small-break tests, a PKL reflood test, and five Marviken critical-flow tests. Analysis efforts in the 2D/3D project concentrated on detailed investigations of Cylindrical-Core Test Facility (CCTF) Core I tests and calculated flow oscillations in the primary loops of the German pressurized water reactor (PWR). Investigations were completed of PWR transients involving emergency feed-water unavailability. Other Light-Water Reactor (LWR) safety progress included the use of the three-dimensional version of the SALE code to study hot-leg injection into the upper plenum and the effect of guide tube cross section on momentum flux. Efforts in Liquid-Metal-Cooled Fast-Breeder Reactor safety included studying transition-phase phenomena in an SNR-300-type reactor geometry using SIMMER and performing Upper Structure Dynamics experiments to examine rupture disk performance. In High-Temperature Gas-Cooled Reactor (HTGR safety, improvements were made to the Composite High-Temperature Gas-Cooled reactor Analysis Program (CHAP) code, and system transients in the Fort St. Vrain reactor were calculated. Other work in this area included thermal stress analyses of core support block response during fire-water cooldown following a loss-of-forced-circulation accident. Tests were run on steel cylinders to determine the effects of the Area Replacement Method on buckling strength as part of the Structural Margins-to-Failure program. In addition, a literature review was completed of models and experiments to determine damping and stiffness of reinforced concrete structures.

  20. Price-Anderson Nuclear Safety Enforcement Program. 1996 Annual report

    SciTech Connect

    1996-01-01

    This first annual report on DOE`s Price Anderson Amendments Act enforcement program covers the activities, accomplishments, and planning for calendar year 1996. It also includes the infrastructure development activities of 1995. It encompasses the activities of the headquarters` Office of Enforcement in the Office of Environment, Safety and Health (EH) and Investigation and the coordinators and technical advisors in DOE`s Field and Program Offices and other EH Offices. This report includes an overview of the enforcement program; noncompliances, investigations, and enforcement actions; summary of significant enforcement actions; examples where enforcement action was deferred; and changes and improvements to the program.

  1. Fault tree applications within the safety program of Idaho Nuclear Corporation

    NASA Technical Reports Server (NTRS)

    Vesely, W. E.

    1971-01-01

    Computerized fault tree analyses are used to obtain both qualitative and quantitative information about the safety and reliability of an electrical control system that shuts the reactor down when certain safety criteria are exceeded, in the design of a nuclear plant protection system, and in an investigation of a backup emergency system for reactor shutdown. The fault tree yields the modes by which the system failure or accident will occur, the most critical failure or accident causing areas, detailed failure probabilities, and the response of safety or reliability to design modifications and maintenance schemes.

  2. Fault tree applications within the safety program of Idaho Nuclear Corporation

    NASA Technical Reports Server (NTRS)

    Vesely, W. E.

    1971-01-01

    Computerized fault tree analyses are used to obtain both qualitative and quantitative information about the safety and reliability of an electrical control system that shuts the reactor down when certain safety criteria are exceeded, in the design of a nuclear plant protection system, and in an investigation of a backup emergency system for reactor shutdown. The fault tree yields the modes by which the system failure or accident will occur, the most critical failure or accident causing areas, detailed failure probabilities, and the response of safety or reliability to design modifications and maintenance schemes.

  3. Educating Next Generation Nuclear Criticality Safety Engineers at the Idaho National Laboratory

    SciTech Connect

    J. D. Bess; J. B. Briggs; A. S. Garcia

    2011-09-01

    One of the challenges in educating our next generation of nuclear safety engineers is the limitation of opportunities to receive significant experience or hands-on training prior to graduation. Such training is generally restricted to on-the-job-training before this new engineering workforce can adequately provide assessment of nuclear systems and establish safety guidelines. Participation in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) can provide students and young professionals the opportunity to gain experience and enhance critical engineering skills. The ICSBEP and IRPhEP publish annual handbooks that contain evaluations of experiments along with summarized experimental data and peer-reviewed benchmark specifications to support the validation of neutronics codes, nuclear cross-section data, and the validation of reactor designs. Participation in the benchmark process not only benefits those who use these Handbooks within the international community, but provides the individual with opportunities for professional development, networking with an international community of experts, and valuable experience to be used in future employment. Traditionally students have participated in benchmarking activities via internships at national laboratories, universities, or companies involved with the ICSBEP and IRPhEP programs. Additional programs have been developed to facilitate the nuclear education of students while participating in the benchmark projects. These programs include coordination with the Center for Space Nuclear Research (CSNR) Next Degree Program, the Collaboration with the Department of Energy Idaho Operations Office to train nuclear and criticality safety engineers, and student evaluations as the basis for their Master's thesis in nuclear engineering.

  4. Safety aspects of ground testing for large nuclear rockets

    SciTech Connect

    Goldman, M.I.

    1988-02-01

    Present nuclear rocket reactors under test in Nevada are operated at nominal power levels of 1000 Mw. It does not seem unreasonable in the future to anticipate reactors with power levels in the range up to 5,000 Mw for space applications. It has been shown that the normal testing of large nuclear rocket engines at NRDS could impose some restrictions on the fuel performance which would not otherwise be required by space flight operation. The only apparent alternative would require a capability for decontaminating effluent gases prior to release to the atmosphere. In addition to the source restrictions, tests will almost certainly be controlled by wind and atmospheric stability conditions, and the requirements for monitoring and control of off-site exposures will be much more stringent than those presently in force. An analysis of maximum accidents indicates that projections of present credible occurrences cannot be tolerated in larger engine tests. The apparent alternatives to a significant (order of magnitude or better) reduction in credible accident consequences, are the establishment of an underground test facility, a facility in an area equivalent to the Pacific weapons proving ground, or in space.

  5. Heterogeneity in utilization of N-glycosylation sites Asn624 and Asn138 in human lactoferrin: a study with glycosylation-site mutants.

    PubMed

    van Berkel, P H; van Veen, H A; Geerts, M E; de Boer, H A; Nuijens, J H

    1996-10-01

    Human lactoferrin (hLF) is a glycoprotein involved in the host defence against infection and excessive inflammation. Our objective was to determine to what extent each of the three sequons for N-linked glycosylation in hLF is actually used. Human kidney-derived 293(S) cell lines expressing recombinant hLF (rhLF) or glycosylation-site mutants were produced. The mutations involved replacement of asparagine residues with glutamine at one or more sequons for N-glycosylation (Asn138, Asn479 and Asn624). Comparative SDS/PAGE analyses of rhLF, mutated rhLF and human-milk-derived (natural) hLF led us to propose that glycosylation of hLF occurs at two sites (at Asn138 and Asn479) in approx. 85% of all hLF molecules. Glycosylation at a single site (Asn479) or at all three sites occurs in approx, 5% and 9% of hLF respectively. The extent of glycosylation at Asn624 was increased to approx. 29% and 40% of Asn479 and Asn138/479 mutant molecules respectively, which indicates that glycosylation at Asn624 in natural hLF might be limited by glycosylation at Asn479. The presence in supernatant of unglycosylated hLF (approx. 60% of the total) after mutations of Asn138 and Asn479 suggests that glycosylation of hLF is not an absolute requirement for its secretion. The pronounced degradation of unglycosylated hLF in supernatant after mutation at all three glycosylation sites (Asn138/479/624 mutant) but not after mutation at both Asn138 and Asn479 suggests that an altered conformation rather than the lack of glycosylation has rendered the Asn138/479/624 mutant susceptible to intra- and/or extra-cellular degradation.

  6. 75 FR 9196 - Letter From Secretary of Energy Accepting Defense Nuclear Facilities Safety Board (Board...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-01

    ... sources; Develop conceptual designs for potential seismic upgrades to key active confinement ventilation... the enclosure to this letter. As noted above, the changes to the DSA currently under review would... DOE directs changes to the nuclear safety posture of its facilities. DOE is expediting its review...

  7. Next Generation Nuclear Plant Structures, Systems, and Components Safety Classification White Paper

    SciTech Connect

    Pete Jordan

    2010-09-01

    This white paper outlines the relevant regulatory policy and guidance for a risk-informed approach for establishing the safety classification of Structures, Systems, and Components (SSCs) for the Next Generation Nuclear Plant and sets forth certain facts for review and discussion in order facilitate an effective submittal leading to an NGNP Combined Operating License application under 10 CFR 52.

  8. Guidelines for Reviewers and the Editor at the Nuclear Safety Information Center.

    ERIC Educational Resources Information Center

    Whetsel, H. B.

    The main purpose of this report is to help novice reviewers accelerate their apprenticeship at the Nuclear Safety Information Center, a computerized information service sponsored by the U.S. Atomic Energy Commission. Guidelines for reviewers are presented in Part 1; Part 2 contains guidelines for the novice editor. The goal of the reviewers and…

  9. Safety Aspects of Nuclear Desalination with Innovative Systems; the EURODESAL Project

    SciTech Connect

    Alessandroni, C.; Cinotti, L.; Mini, G.; Nisan, S.

    2002-07-01

    The proposed paper reports the results of a preliminary investigation on safety impact deriving from the coupling of a desalination plant with a 600 MWe Passive Design PWR like the AP600 Nuclear Power Plant. This evaluation was performed in the frame of the EURODESAL Project of the 5. EURATOM Framework Programme. (authors)

  10. Nuclear safety: Volume 37, Number 3. Technical progress journal, July--September 1996

    SciTech Connect

    Muhlheim, M D

    1996-09-01

    The seven articles in this journal are divided into the following areas: general safety considerations; design features; operating experiences; U.S. Nuclear Regulatory Commission information and analysis; and recent developments. Proposed rule changes since June 30, 1996 are also included.

  11. Index to Nuclear Safety: a technical progress review by chrology, permuted title, and author, Volume 11(1) through Volume 20(6)

    SciTech Connect

    Cottrell, W B; Passiakos, M

    1980-06-01

    This index to Nuclear Safety, a bimonthly technical progress review, covers articles published in Nuclear Safety, Volume II, No. 1 (January-February 1970), through Volume 20, No. 6 (November-December 1979). It is divided into three sections: a chronological list of articles (including abstracts) followed by a permuted-title (KWIC) index and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center (NSIC), covers all safety aspects of nuclear power reactors and associated facilities. Over 600 technical articles published in Nuclear Safety in the last ten years are listed in this index.

  12. Index to Nuclear Safety: a technical progress review by chronology, permuted title, and author, Volume 18 (1) through Volume 22 (6)

    SciTech Connect

    Cottrell, W.B.; Passiakos, M.

    1982-06-01

    This index to Nuclear Safety covers articles published in Nuclear Safety, Volume 18, Number 1 (January-February 1977) through Volume 22, Number 6 (November-December 1981). The index is divided into three section: a chronological list of articles (including abstracts), a permuted-title (KWIC) index, and an author index. Nuclear Safety, a bimonthly technical progress review prepared by the Nuclear Safety Information Center, covers all safety aspects of nuclear power reactors and associated facilities. Over 300 technical articles published in Nuclear Safety in the last 5 years are listed in this index.

  13. 76 FR 13397 - DOE Response to Recommendation 2010-2 of the Defense Nuclear Facilities Safety Board, Pulse Jet...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-11

    ... Response to Recommendation 2010-2 of the Defense Nuclear Facilities Safety Board, Pulse Jet Mixing at the... Defense Nuclear Facilities Safety Board Recommendation 2010-2, concerning Pulse Jet Mixing at the Waste... Board (Board) Recommendation 2010-2, Pulse Jet Mixing at the Waste Treatment and Immobilization Plant...

  14. 78 FR 47011 - Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-02

    ... COMMISSION Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants..., ``Software Unit Testing for Digital Computer Software Used in Safety Systems of Nuclear Power Plants.'' This... software elements if those systems include software. This RG is one of six RG revisions addressing computer...

  15. Current global and Korean issues in radiation safety of nuclear medicine procedures.

    PubMed

    Song, H C

    2016-06-01

    In recent years, the management of patient doses in medical imaging has evolved as concern about radiation exposure has increased. Efforts and techniques to reduce radiation doses are focussed not only on the basis of patient safety, but also on the fundamentals of justification and optimisation in cooperation with international organisations such as the International Commission on Radiological Protection, the International Atomic Energy Agency, and the World Health Organization. The Image Gently campaign in children and Image Wisely campaign in adults to lower radiation doses have been initiated in the USA. The European Association of Nuclear Medicine paediatric dosage card, North American consensus guidelines, and Nuclear Medicine Global Initiative have recommended the activities of radiopharmaceuticals that should be administered in children. Diagnostic reference levels (DRLs), developed predominantly in Europe, may be an important tool to manage patient doses. In Korea, overexposure to radiation, even from the use of medical imaging, has become a public issue, particularly since the accident at the Fukushima nuclear power plant. As a result, the Korean Nuclear Safety and Security Commission revised the technical standards for radiation safety management in medical fields. In parallel, DRLs for nuclear medicine procedures have been collected on a nationwide scale. Notice of total effective dose from positron emission tomography-computed tomography for cancer screening has been mandatory since mid-November 2014. © The International Society for Prosthetics and Orthotics.

  16. 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.

  17. Technical Support Section Instrument Support Program for Nuclear and Nonnuclear Facilities with Safety Requirements

    SciTech Connect

    Adkisson, B.P.

    1995-01-01

    This document describes the requirements, procedures, and responsibilities of the Instrumentation and Controls (I and C) Division's Technical Support Section (TSS) for instruments identified in nonreactor nuclear and nonnuclear facilities at Oak Ridge National Laboratory (ORNL) with Operational Safety Requirements (OSRs) or Limiting Conditions Documents (LCDs). As a result of DOE order 5480.22 Technical Safety Requirements (TSRs), OSRs, and LCDs for nuclear facilities will be eventually replaced by TSRs. OSRs or LCDs will continue to be required for high-, moderate-, or low-level radiological nonnuclear facilities. The objective of this document is to present an instrument surveillance plan for nonreactor nuclear and nonnuclear facility-identified instruments or systems as specified in the facility's OSR, LCD, or TSR. The instrument surveillance plan is a collaborative effort between the facility manager and the I and C Division TSS staff, thereby ensuring that the surveillance requirements stated in the OSR, LCD, or TSR are fulfilled within the required time frame.

  18. Optimization approach for evaluation of allowed outage times in nuclear-safety systems. [PWR; BWR

    SciTech Connect

    Farahzad, P.

    1983-01-01

    The purpose of this paper is to develop and demonstrate an approach for determining allowed outage times (AOTs) of nuclear systems based on linear programming techniques. Presently nuclear power plants are operated within the constraints of technical specifications defined by the Nuclear Regulatory Commission. These specifications, among other things, define the time a safety system component may be allowed to be serviced for repair without bringing the plant to hot shutdown condition. The time the component is allowed to be serviced is commonly known as the allowed outage time and the determination of such times is presently based on engineering judgements. Over the last few years, efforts were made to develop allowed outage times for safety system components based on probabilistic considerations. The method given here is based on linear programming and it provides a tool for simultaneous consideration and evaluation of any number of linear constraints imposed on the problem.

  19. Discussion on software aging management of nuclear power plant safety digital control system.

    PubMed

    Liang, Huihui; Gu, Pengfei; Tang, Jianzhong; Chen, Weihua; Gao, Feng

    2016-01-01

    Managing the aging of digital control systems ensures that nuclear power plant systems are in adequate safety margins during their life cycles. Software is a core component in the execution of control logic and differs between digital and analog control systems. The hardware aging management for the digital control system is similar to that for the analog system, which has matured over decades of study. However, software aging management is still in the exploratory stage. Software aging evaluation is critical given the higher reliability and safety requirements of nuclear power plants. To ensure effective inputs for reliability assessment, this paper provides the required software aging information during the life cycle. Moreover, the software aging management scheme for safety digital control system is proposed on the basis of collected aging information.

  20. Climate considerations in long-term safety assessments for nuclear waste repositories.

    PubMed

    Näslund, Jens-Ove; Brandefelt, Jenny; Liljedahl, Lillemor Claesson

    2013-05-01

    For a deep geological repository for spent nuclear fuel planned in Sweden, the safety assessment covers up to 1 million years. Climate scenarios range from high-end global warming for the coming 100 000 years, through deep permafrost, to large ice sheets during glacial conditions. In contrast, in an existing repository for short-lived waste the activity decays to low levels within a few tens of thousands of years. The shorter assessment period, 100 000 years, requires more focus on climate development over the coming tens of thousands of years, including the earliest possibility for permafrost growth and freezing of the engineered system. The handling of climate and climate change in safety assessments must be tailor-made for each repository concept and waste type. However, due to the uncertain future climate development on these vast time scales, all safety assessments for nuclear waste repositories require a range of possible climate scenarios.

  1. Organizational analysis and safety for utilities with nuclear power plants: an organizational overview. Volume 1. [PWR; BWR

    SciTech Connect

    Osborn, R.N.; Olson, J.; Sommers, P.E.; McLaughlin, S.D.; Jackson, M.S.; Scott, W.G.; Connor, P.E.

    1983-08-01

    This two-volume report presents the results of initial research on the feasibility of applying organizational factors in nuclear power plant (NPP) safety assessment. A model is introduced for the purposes of organizing the literature review and showing key relationships among identified organizational factors and nuclear power plant safety. Volume I of this report contains an overview of the literature, a discussion of available safety indicators, and a series of recommendations for more systematically incorporating organizational analysis into investigations of nuclear power plant safety.

  2. Investigation of criticality safety control infraction data at a nuclear facility

    SciTech Connect

    Cournoyer, Michael E.; Merhege, James F.; Costa, David A.; Art, Blair M.; Gubernatis, David C.

    2014-10-27

    Chemical and metallurgical operations involving plutonium and other nuclear materials account for most activities performed at the LANL's Plutonium Facility (PF-4). The presence of large quantities of fissile materials in numerous forms at PF-4 makes it necessary to maintain an active criticality safety program. The LANL Nuclear Criticality Safety (NCS) Program provides guidance to enable efficient operations while ensuring prevention of criticality accidents in the handling, storing, processing and transportation of fissionable material at PF-4. In order to achieve and sustain lower criticality safety control infraction (CSCI) rates, PF-4 operations are continuously improved, through the use of Lean Manufacturing and Six Sigma (LSS) business practices. Employing LSS, statistically significant variations (trends) can be identified in PF-4 CSCI reports. In this study, trends have been identified in the NCS Program using the NCS Database. An output metric has been developed that measures ADPSM Management progress toward meeting its NCS objectives and goals. Using a Pareto Chart, the primary CSCI attributes have been determined in order of those requiring the most management support. Data generated from analysis of CSCI data help identify and reduce number of corresponding attributes. In-field monitoring of CSCI's contribute to an organization's scientific and technological excellence by providing information that can be used to improve criticality safety operation safety. This increases technical knowledge and augments operational safety.

  3. Investigation of criticality safety control infraction data at a nuclear facility

    DOE PAGES

    Cournoyer, Michael E.; Merhege, James F.; Costa, David A.; ...

    2014-10-27

    Chemical and metallurgical operations involving plutonium and other nuclear materials account for most activities performed at the LANL's Plutonium Facility (PF-4). The presence of large quantities of fissile materials in numerous forms at PF-4 makes it necessary to maintain an active criticality safety program. The LANL Nuclear Criticality Safety (NCS) Program provides guidance to enable efficient operations while ensuring prevention of criticality accidents in the handling, storing, processing and transportation of fissionable material at PF-4. In order to achieve and sustain lower criticality safety control infraction (CSCI) rates, PF-4 operations are continuously improved, through the use of Lean Manufacturing andmore » Six Sigma (LSS) business practices. Employing LSS, statistically significant variations (trends) can be identified in PF-4 CSCI reports. In this study, trends have been identified in the NCS Program using the NCS Database. An output metric has been developed that measures ADPSM Management progress toward meeting its NCS objectives and goals. Using a Pareto Chart, the primary CSCI attributes have been determined in order of those requiring the most management support. Data generated from analysis of CSCI data help identify and reduce number of corresponding attributes. In-field monitoring of CSCI's contribute to an organization's scientific and technological excellence by providing information that can be used to improve criticality safety operation safety. This increases technical knowledge and augments operational safety.« less

  4. SRTC criticality safety technical review: Nuclear criticality safety evaluation 94-02, uranium solidification facility pencil tank module spacing

    SciTech Connect

    Rathbun, R.

    1994-04-26

    Review of NMP-NCS-94-0087, ``Nuclear Criticality Safety Evaluation 94-02: Uranium Solidification Facility Pencil Tank Module Spacing (U), April 18, 1994,`` was requested of the SRTC Applied Physics Group. The NCSE is a criticality assessment to show that the USF process module spacing, as given in Non-Conformance Report SHM-0045, remains safe for operation. The NCSE under review concludes that the module spacing as given in Non-Conformance Report SHM-0045 remains in a critically safe configuration for all normal and single credible abnormal conditions. After a thorough review of the NCSE, this reviewer agrees with that conclusion.

  5. 75 FR 52046 - Development of U.S. Nuclear Regulatory Commission Safety Culture Policy Statement: Public Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-24

    ... COMMISSION Development of U.S. Nuclear Regulatory Commission Safety Culture Policy Statement: Public Meeting... solicit comments on the revision of its draft safety culture policy statement, including the revised...; ML093030375), the results of the NRC's February 2010 workshop (February workshop) on safety culture, and...

  6. The Fukushima Dai-ichi Accident and its implications for the safety of nuclear power

    NASA Astrophysics Data System (ADS)

    Barletta, William

    2016-05-01

    Five years ago the dramatic events in Fukushima that followed the massive earthquake and subsequent tsunami that struck Japan on March 11, 2011 sharpened the focus of scientists, engineers and general public on the broad range of technical, environmental and societal issues involved in assuring the safety of the world's nuclear power complex. They also called into question the potential of nuclear power to provide a growing, sustainable resource of CO2-free energy. The issues raised by Fukushima Dai-ichi have provoked urgent concern, not only because of the potential harm that could result from severe accidents or from intentional damage to nuclear reactors or to facilities involved in the nuclear fuel cycle, but also because of the extensive economic impact of those accidents and of the measures taken to avoid them.

  7. Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

    SciTech Connect

    1993-09-01

    The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses.

  8. Commercial grade item (CGI) dedication of MDR relays for nuclear safety related applications

    NASA Astrophysics Data System (ADS)

    Das, Ranjit K.; Julka, Anil; Modi, Govind

    1994-08-01

    MDR relays manufactured by Potter & Brumfield (P&B) have been used in various safety related applications in commercial nuclear power plants. These include emergency safety features (ESF) actuation systems, emergency core cooling systems (ECCS) actuation, and reactor protection systems. The MDR relays manufactured prior to May 1990 showed signs of generic failure due to corrosion and outgassing of coil varnish. P&B has made design changes to correct these problems in relays manufactured after May 1990. However, P&B does not manufacture the relays under any 10CFR50 Appendix B quality assurance (QA) program. They manufacture the relays under their commercial QA program and supply these as commercial grade items. This necessitates CGI Dedication of these relays for use in nuclear-safety-related applications. This paper presents a CGI dedication program that has been used to dedicate the MDR relays manufactured after been used to dedicate the MDR relays manufactured after May 1990. The program is in compliance with current Nuclear Regulatory Commission (NRC) and Electric Power Research Institute (EPRI) guidelines and applicable industry standards; it specifies the critical characteristics of the relays, provides the tests and analysis required to verify the critical characteristics, the acceptance criteria for the test results, performs source verification to quality P&B for its control of the critical characteristics, and provides documentation. The program provides reasonable assurance that the new MDR relays will perform their intended safety functions.

  9. ORNL Nuclear Safety Research and Development Program Bimonthly Report for July-August 1968

    SciTech Connect

    Cottrell, W.B.

    2001-08-17

    The accomplishments during the months of July and August in the research and development program under way at ORNL as part of the U.S. Atomic Energy Commission's Nuclear Safety Program are summarized, Included in this report are work on various chemical reactions, as well as the release, characterization, and transport of fission products in containment systems under various accident conditions and on problems associated with the removal of these fission products from gas streams. Although most of this work is in general support of water-cooled power reactor technology, including LOFT and CSE programs, the work reflects the current safety problems, such as measurements of the prompt fuel element failure phenomena and the efficacy of containment spray and pool-suppression systems for fission-product removal. Several projects are also conducted in support of the high-temperature gas-cooled reactor (HTGR). Other major projects include fuel-transport safety investigations, a series of discussion papers on various aspects of water-reactor technology, antiseismic design of nuclear facilities, and studies of primary piping and steel, pressure-vessel technology. Experimental work relative to pressure-vessel technology includes investigations of the attachment of nozzles to shells and the implementation of joint AEX-PVFX programs on heavy-section steel technology and nuclear piping, pumps, and valves. Several of the projects are directly related to another major undertaking; namely, the AEC's standards program, which entails development of engineering safeguards and the establishment of codes and standards for government-owned or -sponsored reactor facilities. Another task, CHORD-S, is concerned with the establishment of computer programs for the evaluation of reactor design data, The recent activities of the NSIC and the Nuclear Safety journal in behalf of the nuclear community are also discussed.

  10. ORNL Nuclear Safety Research and Development Program Bimonthly Report for September-October 1968

    SciTech Connect

    Cottrell, W.B.

    2001-08-17

    The accomplishments during the months of September and October in the research and development program under way at ORNL as part of the U.S. Atomic Energy Commission's Nuclear Safety Program are summarized. Included in this report are work on various chemical reactions, as well as the release, characterization, and transport of fission products in containment systems under various accident conditions and on problems associated with the removal of these fission products from gas streams. Although most of this work is in general support of water-cooled power reactor technology, including LOFT and CSE programs, the work reflects the current safety problems, such as measurements of the prompt fuel element failure phenomena and the efficacy of containment spray and pool-suppression systems for fission-product removal. Several projects are also conducted in support of the high-temperature gas-cooled reactor (HTGR). Other major projects include fuel-transport safety investigations, a series cf discussion papers on various aspects of water-reactor technology, antiseismic design of nuclear facilities, and studies of primary piping and steel pressure-vessel technology. Experimental work relative to pressure-vessel technology includes investigations of the attachment of nozzles to shells and the implementation of joint AEC-PVRC programs on heavy-section steel technology and nuclear piping, pumps, and valves. Several of the projects are directly related to another major undertaking; namely, the AEC's standards program, which entails development of engineering safeguards and the establishment of codes and standards for government-owned or -sponsored reactor facilities. Another task, CHORD-S, is concerned with the establishment of computer programs for the evaluation of reactor design data. The recent activities of the NSIC and the Nuclear Safety journal in behalf of the nuclear community are also discussed.

  11. 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.

  12. Panel session on "safety, health and the environment: implications of nuclear power growth".

    PubMed

    Bilbao y León, Sama

    2011-01-01

    This paper summarizes the presentations and the insights offered by panelists John P. Winston, Robert Bernero, and Stephen LaMontagne during the Panel on Safety, Health and the Environment: Implications of Nuclear Power Growth that took place during the NCRP 2009 Annual Meeting. The paper describes the opportunities and the challenges faced in the areas of infrastructure development, radiation control, licensing and regulatory issues, and non-proliferation as a consequence of the forecasted growth in nuclear power capacity worldwide. Copyright © 2010 Health Physics Society

  13. Sensitivity and Uncertainty Analysis for Nuclear Criticality Safety Using KENO in the SCALE Code System

    NASA Astrophysics Data System (ADS)

    Rearden, B. T.

    Sensitivity and uncertainty methods have been developed to aid in the establishment of areas of applicability and validation of computer codes and nuclear data for nuclear criticality safety studies. A key component in this work is the generation of sensitivity and uncertainty parameters for typically several hundred benchmarks experiments used in validation exercises. Previously, only one-dimensional sensitivity tools were available for this task, which necessitated the remodeling of multidimensional inputs in order for such an analysis to be performed. This paper describes the development of the SEN3 Monte Carlo based sensitivity analysis sequence for SCALE.

  14. Increasing efficiency safety of cooling systems in a floating nuclear power plant

    NASA Astrophysics Data System (ADS)

    Yenivatov, V.; Fedorovsky, K.; Zhelezniak, A.; Bordyug, A.

    2017-01-01

    In order to increase the reliability and safety of a floating nuclear power plant, the research proposes to apply closed-loop cooling systems for floating nuclear power plants’ cooling. A ship hull is proposed to be used as a surface of the heat exchanger, known as a ship hull heat exchanger. This paper presents detailed study on the effect of the ship hull angle on the heat transfer coefficient and measures to improve the heat transfer coefficient. It is concluded that application of gas-liquid flows to the seaside surfaces of the ship hull is one of the most effective ways of improving the heat transfer coefficient.

  15. Nuclear criticality safety assessment of the low level radioactive waste disposal facility trenches

    SciTech Connect

    Kahook, S.D.

    1994-04-01

    Results of the analyses performed to evaluate the possibility of nuclear criticality in the Low Level Radioactive Waste Disposal Facility (LLRWDF) trenches are documented in this report. The studies presented in this document are limited to assessment of the possibility of criticality due to existing conditions in the LLRWDF. This document does not propose nor set limits for enriched uranium (EU) burial in the LLRWDF and is not a nuclear criticality safety evaluation nor analysis. The calculations presented in the report are Level 2 calculations as defined by the E7 Procedure 2.31, Engineering Calculations.

  16. General-purpose heat source project and space nuclear safety fuels program. Progress report, February 1980

    SciTech Connect

    Maraman, W.J.

    1980-05-01

    This formal monthly report covers the studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of the Los Alamos Scientific Laboratory. The two programs involved are: General-Purpose Heat Source Development and Space Nuclear Safety and Fuels. Most of the studies discussed here are of a continuing nature. Results and conclusions described may change as the work continues. Published reference to the results cited in this report should not be made without the explicit permission of the person in charge of the work.

  17. Vulnerability, safety and response of nuclear power plants to the hydroclimatic hazards

    NASA Astrophysics Data System (ADS)

    János Katona, Tamás; Vilimi, András

    2016-04-01

    The Great Tohoku Earthquake and Tsunami, and the severe accident at Fukushima Dai-ichi nuclear power plant 2011 alerted the nuclear industry to danger of extreme rare natural hazards. The subsequent "stress tests" performed by the nuclear industry in Europe and all over the world identifies the nuclear power plant (NPP) vulnerabilities and define the measures for increasing the plant safety. According to the international practice of nuclear safety regulations, the cumulative core damage frequency for NPPs has to be 10-5/a, and the cumulative frequency of early large release has to be 10-6/a. In case of operating plants these annual probabilities can be little higher, but the licensees are obliged to implement all reasonable practicable measures for increasing the plant safety. For achieving the required level of safety, design basis of NPPs for natural hazards has to be defined at the 10-4/a ⎯10-5/a levels of annual exceedance probability. Tornado hazard is some kind of exception, e.g., the design basis annual probability for tornado in the US is equal to 10-7/a. Design of the NPPs shall provide for an adequate margin to protect items ultimately necessary to prevent large or early radioactive releases in the event of levels of natural hazards exceeding those to be considered for design. The plant safety has to be reviewed for accounting the changes of the environmental conditions and natural hazards in case of necessity, but as minimum every ten years in the frame of periodic safety reviews. Long-term forecast of environmental conditions and hazards has to be accounted for in the design basis of the new plants. Changes in hydroclimatic variables, e.g., storms, tornadoes, river floods, flash floods, extreme temperatures, droughts affect the operability and efficiency as well as the safety the NPPs. Low flow rates and high water temperature in the rivers may force to operate at reduced power level or shutdown the plant (Cernavoda NPP, Romania, August 2009). The

  18. Technical Guidance from the International Safety Framework for Nuclear Power Source Applications in Outer Space for Design and Development Phases

    NASA Astrophysics Data System (ADS)

    Summerer, Leopold

    2014-08-01

    In 2009, the International Safety Framework for Nuclear Power Source Applications in Outer Space [1] has been adopted, following a multi-year process that involved all major space faring nations in the frame of the International Atomic Energy Agency and the UN Committee on the Peaceful Uses of Outer Space. The safety framework reflects an international consensus on best practices. After the older 1992 Principles Relevant to the Use of Nuclear Power Sources in Outer Space, it is the second document at UN level dedicated entirely to space nuclear power sources.This paper analyses aspects of the safety framework relevant for the design and development phases of space nuclear power sources. While early publications have started analysing the legal aspects of the safety framework, its technical guidance has not yet been subject to scholarly articles. The present paper therefore focuses on the technical guidance provided in the safety framework, in an attempt to assist engineers and practitioners to benefit from these.

  19. Nuclear electric propulsion operational reliability and crew safety study: NEP systems/modeling report

    NASA Technical Reports Server (NTRS)

    Karns, James

    1993-01-01

    The objective of this study was to establish the initial quantitative reliability bounds for nuclear electric propulsion systems in a manned Mars mission required to ensure crew safety and mission success. Finding the reliability bounds involves balancing top-down (mission driven) requirements and bottom-up (technology driven) capabilities. In seeking this balance we hope to accomplish the following: (1) provide design insights into the achievability of the baseline design in terms of reliability requirements, given the existing technology base; (2) suggest alternative design approaches which might enhance reliability and crew safety; and (3) indicate what technology areas require significant research and development to achieve the reliability objectives.

  20. Scenario Analysis for the Safety Assessment of Nuclear Waste Repositories: A Critical Review.

    PubMed

    Tosoni, Edoardo; Salo, Ahti; Zio, Enrico

    2017-09-05

    A major challenge in scenario analysis for the safety assessment of nuclear waste repositories pertains to the comprehensiveness of the set of scenarios selected for assessing the safety of the repository. Motivated by this challenge, we discuss the aspects of scenario analysis relevant to comprehensiveness. Specifically, we note that (1) it is necessary to make it clear why scenarios usually focus on a restricted set of features, events, and processes; (2) there is not yet consensus on the interpretation of comprehensiveness for guiding the generation of scenarios; and (3) there is a need for sound approaches to the treatment of epistemic uncertainties. © 2017 Society for Risk Analysis.

  1. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    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 EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  2. Annual Report To Congress. Department of Energy Activities Relating to the Defense Nuclear Facilities Safety Board, Calendar Year 2003

    SciTech Connect

    None, None

    2004-02-28

    The Department of Energy (Department) submits an Annual Report to Congress each year detailing the Department’s activities relating to the Defense Nuclear Facilities Safety Board (Board), which provides advice and recommendations to the Secretary of Energy (Secretary) regarding public health and safety issues at the Department’s defense nuclear facilities. In 2003, the Department continued ongoing activities to resolve issues identified by the Board in formal recommendations and correspondence, staff issue reports pertaining to Department facilities, and public meetings and briefings. Additionally, the Department is implementing several key safety initiatives to address and prevent safety issues: safety culture and review of the Columbia accident investigation; risk reduction through stabilization of excess nuclear materials; the Facility Representative Program; independent oversight and performance assurance; the Federal Technical Capability Program (FTCP); executive safety initiatives; and quality assurance activities. The following summarizes the key activities addressed in this Annual Report.

  3. Just in Time DSA-The Hanford Nuclear Safety Basis Strategy

    SciTech Connect

    Olinger, S. J.; Buhl, A. R.

    2002-02-26

    The U.S. Department of Energy, Richland Operations Office (RL) is responsible for 30 hazard category 2 and 3 nuclear facilities that are operated by its prime contractors, Fluor Hanford Incorporated (FHI), Bechtel Hanford, Incorporated (BHI) and Pacific Northwest National Laboratory (PNNL). The publication of Title 10, Code of Federal Regulations, Part 830, Subpart B, Safety Basis Requirements (the Rule) in January 2001 imposed the requirement that the Documented Safety Analyses (DSA) for these facilities be reviewed against the requirements of the Rule. Those DSA that do not meet the requirements must either be upgraded to satisfy the Rule, or an exemption must be obtained. RL and its prime contractors have developed a Nuclear Safety Strategy that provides a comprehensive approach for supporting RL's efforts to meet its long term objectives for hazard category 2 and 3 facilities while also meeting the requirements of the Rule. This approach will result in a reduction of the total number of safety basis documents that must be developed and maintained to support the remaining mission and closure of the Hanford Site and ensure that the documentation that must be developed will support: compliance with the Rule; a ''Just-In-Time'' approach to development of Rule-compliant safety bases supported by temporary exemptions; and consolidation of safety basis documents that support multiple facilities with a common mission (e.g. decontamination, decommissioning and demolition [DD&D], waste management, surveillance and maintenance). This strategy provides a clear path to transition the safety bases for the various Hanford facilities from support of operation and stabilization missions through DD&D to accelerate closure. This ''Just-In-Time'' Strategy can also be tailored for other DOE Sites, creating the potential for large cost savings and schedule reductions throughout the DOE complex.

  4. A Logical Approach to Designing Safety Test Plans for Space Nuclear Systems

    SciTech Connect

    Coleman, James R

    2004-02-04

    This paper presents a logical approach to designing a safety test plan for a space nuclear system. It is pointed out that two important facts need to underlie the development of a test plan: first, that sequential insults and the accumulation of damage are the rule; and second that the response of the nuclear system is stochastic (i.e., for any given set of conditions a probabilistic range of outcomes will occur regardless of the state of our knowledge). Because of these facts a deterministic approach can only be a starting point. The substance of the approach consists of undertaking and documenting three basic efforts: (1) a description of the analysts view of the problem and how it fits into the safety analysis, (2) a formal documentation of the purpose and requirements of the test plan (or test), and (3) an assessment of the use or usefulness of existing test data.

  5. Handbook of nuclear power plant seismic fragilities, Seismic Safety Margins Research Program

    SciTech Connect

    Cover, L.E.; Bohn, M.P.; Campbell, R.D.; Wesley, D.A.

    1983-12-01

    The Seismic Safety Margins Research Program (SSMRP) has a gola to develop a complete fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. As part of this program, calculations of the seismic risk from a typical commercial nuclear reactor were made. These calculations required a knowledge of the probability of failure (fragility) of safety-related components in the reactor system which actively participate in the hypothesized accident scenarios. This report describes the development of the required fragility relations and the data sources and data reduction techniques upon which they are based. Both building and component fragilities are covered. The building fragilities are for the Zion Unit 1 reactor which was the specific plant used for development of methodology in the program. Some of the component fragilities are site-specific also, but most would be usable for other sites as well.

  6. Introduction to the nuclear criticality safety evaluation of facility X-705, Portsmouth Gaseous Diffusion Plant

    SciTech Connect

    Sheaffer, M.K.; Keeton, S.C.

    1993-08-16

    This report is the first in a series of documents that will evaluate nuclear criticality safety in the Decontamination and Recovery Facility, X-705, Portsmouth Gaseous Diffusion Plant. It provides an overview of the facility, categorizes its functions for future analysis, reviews existing NCS documentation, and explains the follow-on effort planned for X-705. A detailed breakdown of systems, subsystems, and operational areas is presented and cross-referenced to existing NCS documentation.

  7. Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

    SciTech Connect

    Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; Scaglione, John M.

    2016-06-15

    We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.

  8. Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

    SciTech Connect

    Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; Scaglione, John M.

    2016-06-15

    We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance. These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.

  9. Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks

    DOE PAGES

    Banerjee, Kaushik; Robb, Kevin R.; Radulescu, Georgeta; ...

    2016-06-15

    We completed a novel assessment to determine the unquantified and uncredited safety margins (i.e., the difference between the licensing basis and as-loaded calculations) available in as-loaded spent nuclear fuel (SNF) casks. This assessment was performed as part of a broader effort to assess issues and uncertainties related to the continued safety of casks during extended storage and transportability following extended storage periods. Detailed analyses crediting the actual as-loaded cask inventory were performed for each of the casks at three decommissioned pressurized water reactor (PWR) sites to determine their characteristics relative to regulatory safety criteria for criticality, thermal, and shielding performance.more » These detailed analyses were performed in an automated fashion by employing a comprehensive and integrated data and analysis tool—Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). Calculated uncredited criticality margins from 0.07 to almost 0.30 Δkeff were observed; calculated decay heat margins ranged from 4 to almost 22 kW (as of 2014); and significant uncredited transportation dose rate margins were also observed. The results demonstrate that, at least for the casks analyzed here, significant uncredited safety margins are available that could potentially be used to compensate for SNF assembly and canister structural performance related uncertainties associated with long-term storage and subsequent transportation. The results also suggest that these inherent margins associated with how casks are loaded could support future changes in cask licensing to directly or indirectly credit the margins. Work continues to quantify the uncredited safety margins in the SNF casks loaded at other nuclear reactor sites.« less

  10. International Nuclear Safety Center database on thermophysical properties of reactor materials

    SciTech Connect

    Fink, J.K.; Sofu, T.; Ley, H.

    1997-08-01

    The International Nuclear Safety Center (INSC) database has been established at Argonne National Laboratory to provide easily accessible data and information necessary to perform nuclear safety analyses and to promote international collaboration through the exchange of nuclear safety information. The INSC database, located on the World Wide Web at http://www.insc.anl.gov, contains critically assessed recommendations for reactor material properties for normal operating conditions, transients, and severe accidents. The initial focus of the database is on thermodynamic and transport properties of materials for water reactors. Materials that are being included in the database are fuel, absorbers, cladding, structural materials, coolant, and liquid mixtures of combinations of UO{sub 2}, ZrO{sub 2}, Zr, stainless steel, absorber materials, and concrete. For each property, the database includes: (1) a summary of recommended equations with uncertainties; (2) a detailed data assessment giving the basis for the recommendations, comparisons with experimental data and previous recommendations, and uncertainties; (3) graphs showing recommendations, uncertainties, and comparisons with data and other equations; and (4) property values tabulated as a function of temperature.

  11. Lessons Learned for Space Safety from the Fukushima Nuclear Power Plant Accident

    NASA Astrophysics Data System (ADS)

    Nogami, Manami; Miki, Masami; Mitsui, Masami; Kawada, Ysuhiro; Takeuchi, Nobuo

    2013-09-01

    On March 11 2011, Tohoku Region Pacific Coast Earthquake hit Japan and caused the devastating damage. The Fukushima Nuclear Power Station (NPS) was also severely damaged.The Japanese NPSs are designed based on the detailed safety requirements and have multiple-folds of hazard controls to the catastrophic hazards as in space system. However, according to the initial information from the Tokyo Electric Power Company (TEPCO) and the Japanese government, the larger-than-expected tsunami and subsequent events lost the all hazard controls to the release of radioactive materials.At the 5th IAASS, Lessons Learned from this disaster was reported [1] mainly based on the "Report of the Japanese Government to the IAEA Ministerial Conference on Nuclear Safety" [2] published by Nuclear Emergency Response Headquarters in June 2011, three months after the earthquake.Up to 2012 summer, the major investigation boards, including the Japanese Diet, the Japanese Cabinet and TEPCO, published their final reports, in which detailed causes of this accident and several recommendations are assessed from each perspective.In this paper, the authors examine to introduce the lessons learned to be applied to the space safety as findings from these reports.

  12. Nuclear criticality safety evaluation -- DWPF Late Wash Facility, Salt Process Cell and Chemical Process Cell

    SciTech Connect

    Williamson, T.G.

    1994-10-17

    The Savannah River Site (SRS) High Level Nuclear Waste will be vitrified in the Defense Waste Processing Facility (DWPF) for long term storage and disposal. This is a nuclear criticality safety evaluation for the Late Wash Facility (LWF), the Salt Processing Cell (SPC) and the Chemical Processing Cell (CPC). of the DWPF. Waste salt solution is processed in the Tank Farm In-Tank Precipitation (ITP) process and is then further washed in the DWPF Late Wash Facility (LWF) before it is fed to the DWPF Salt Processing Cell. In the Salt Processing Cell the precipitate slurry is processed in the Precipitate Reactor (PR) and the resultant Precipitate Hydrolysis Aqueous (PHA) produce is combined with the sludge feed and frit in the DWPF Chemical Process Cell to produce a melter feed. The waste is finally immobilized in the Melt Cell. Material in the Tank Farm and the ITP and Extended Sludge processes have been shown to be safe against a nuclear criticality by others. The precipitate slurry feed from ITP and the first six batches of sludge feed are safe against a nuclear criticality and this evaluation demonstrates that the processes in the LWF, the SPC and the CPC do not alter the characteristics of the materials to compromise safety.

  13. 76 FR 16758 - DOE Response to Recommendation 2010-1 of the Defense Nuclear Facilities Safety Board, Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-25

    ...: Ms. Amanda Anderson, Nuclear Engineer, Departmental Representative to the Defense Nuclear Facilities... Security. February 28, 2011 The Honorable Peter S. Winokur Chairman, Defense Nuclear Facilities...

  14. Conduct and results of the Interagency Nuclear Safety Review Panel's evaluation of the Ulysses space mission

    SciTech Connect

    Sholtis, J.A. Jr. ); Gray, L.B. ); Huff, D.A. ); Klug, N.P. ); Winchester, R.O. )

    1991-01-01

    The recent 6 October 1990 launch and deployment of the nuclear-powered Ulysses spacecraft from the Space Shuttle {ital Discovery} culminated an extensive safety review and evaluation effort by the Interagency Nuclear Safety Review Panel (INSRP). After more than a year of detailed independent review, study, and analysis, the INSRP prepared a Safety Evaluation Report (SER) on the Ulysses mission, in accordance with Presidential Directive-National Security Council memorandum 25. The SER, which included a review of the Ulysses Final Safety Analysis Report (FSAR) and an independent characterization of the mission risks, was used by the National Aeronautics and Space Administration (NASA) in its decision to request launch approval as well as by the Executive Office of the President in arriving at a launch decision based on risk-benefit considerations. This paper provides an overview of the Ulysses mission and the conduct as well as the results of the INSRP evaluation. While the mission risk determined by the INSRP in the SER was higher than that characterized by the Ulysses project in the FSAR, both reports indicated that the radiological risks were relatively small. In the final analysis, the SER proved to be supportive of a positive launch decision. The INSRP evaluation process has demonstrated its effectiveness numerous times since the 1960s. In every case, it has provided the essential ingredients and perspective to permit an informed launch decision at the highest level of our Government.

  15. Nuclear power and probabilistic safety assessment (PSA): past through future applications

    NASA Astrophysics Data System (ADS)

    Stamatelatos, M. G.; Moieni, P.; Everline, C. J.

    1995-03-01

    Nuclear power reactor safety in the United States is about to enter a new era -- an era of risk- based management and risk-based regulation. First, there was the age of `prescribed safety assessment,' during which a series of design-basis accidents in eight categories of severity, or classes, were postulated and analyzed. Toward the end of that era, it was recognized that `Class 9,' or `beyond design basis,' accidents would need special attention because of the potentially severe health and financial consequences of these accidents. The accident at Three Mile Island showed that sequences of low-consequence, high-frequency events and human errors can be much more risk dominant than the Class 9 accidents. A different form of safety assessment, PSA, emerged and began to gain ground against the deterministic safety establishment. Eventually, this led to the current regulatory requirements for individual plant examinations (IPEs). The IPEs can serve as a basis for risk-based regulation and management, a concept that may ultimately transform the U.S. regulatory process from its traditional deterministic foundations to a process predicated upon PSA. Beyond the possibility of a regulatory environment predicated upon PSA lies the possibility of using PSA as the foundation for managing daily nuclear power plant operations.

  16. Barriers and solutions in implementing occupational health and safety services at a large nuclear weapons facility.

    PubMed

    Takaro, T K; Ertell, K; Salazar, M K; Beaudet, N; Stover, B; Hagopian, A; Omenn, G; Barnhart, S

    2000-01-01

    The Hanford Nuclear Reservation is one of the U.S. Department of Energy's largest nuclear weapons sites. The enormous changes experienced by Hanford over the last several years, as its mission has shifted from weapons production to cleanup, has profoundly affected its occupational health and safety services. Innovative programs and new initiatives hold promise for a safer workplace for the thousands of workers at Hanford and other DOE sites. However, occupational health and safety professionals continue to face multiple organizational, economic, and cultural challenges. A major problem identified during this review was the lack of coordination of onsite services. Because each health and safety program operates independently (albeit with the guidance of the Richland field operations office), many services are duplicative and the health and safety system is fragmented. The fragmentation is compounded by the lack of centralized data repositories for demographic and exposure data. Innovative measures such as a questionnaire-driven Employee Job Task Analysis linked to medical examinations has allowed the site to move from the inefficient and potentially dangerous administrative medical monitoring assignment to defensible risk-based assignments and could serve as a framework for improving centralized data management and service delivery.

  17. Safety team assessments at NRC (Nuclear Regulatory Commission)-licensed fuel facilities

    SciTech Connect

    Sjoblom, G.L.

    1988-01-01

    Following the hydraulic rupture of a UF cylinder at the Sequoyah Fuels Facility on January 4, 1986, the US Nuclear Regulatory Commission's (NRC's) executive director for operations (EDO) established an augmented inspection team to investigate the accident. The investigation is reported in NUREG-1179. The EDO then formed a lessons-learned group to report on the action NRC might reasonably take to prevent similar accidents. The group's recommendations are reported in NUREG-1198. In addition, the EDO formed an independent materials safety regulation review study group (MSRRSG) to review the licensing and inspection program for NRC-licensed fuel cycle and materials facilities. During the same period of time that the MSRRSG report was being prepared and evaluated, the staff undertook an independent action to assess operational safety at each of the 12 major fuel facilities licensed by the NRC. The facilities included the 2 facilities producing uranium hexafluoride, the 7 facilities producing commercial nuclear reactor fuel, and the 3 facilities producing naval reactor fuel. The most important safety issues identified as needing attention by licensees were in the areas of fire protection, chemical hazards identification and mitigation, management controls or quality assurance, safety-related instrumentation and maintenance, and emergency preparedness.

  18. Robotic and nuclear safety for an automated/teleoperated glove box system

    SciTech Connect

    Domning, E.E. ); McMahon, T.T.; Sievers, R.H. )

    1991-09-01

    Lawrence Livermore National Laboratory (LLNL) is developing a fully automated system to handle the processing of special nuclear materials (SNM). This work is performed in response to the new goals at the Department of Energy (DOE) for hazardous waste minimization and radiation dose reduction. This fully automated system, called the automated test bed (ATB), consists of an IBM gantry robot and automated processing equipment sealed within a glove box. While the ATB is a cold system, we are designing it as a prototype of the future hot system. We recognized that identification and application of safety requirements early in the design phase will lead to timely installation and approval of the hot system. This paper identifies these safety issues as well as the general safety requirements necessary for the safe operation of the ATB. 4 refs., 2 figs.

  19. The International Safety Framework for nuclear power source applications in outer space-Useful and substantial guidance

    NASA Astrophysics Data System (ADS)

    Summerer, L.; Wilcox, R. E.; Bechtel, R.; Harbison, S.

    2015-06-01

    In 2009, the International Safety Framework for Nuclear Power Source Applications in Outer Space was adopted, following a multi-year process that involved all major space faring nations under the auspices of a partnership between the UN Committee on the Peaceful Uses of Outer Space and the International Atomic Energy Agency. The Safety Framework reflects an international consensus on best practices to achieve safety. Following the 1992 UN Principles Relevant to the Use of Nuclear Power Sources in Outer Space, it is the second attempt by the international community to draft guidance promoting the safety of applications of nuclear power sources in space missions. NPS applications in space have unique safety considerations compared with terrestrial applications. Mission launch and outer space operational requirements impose size, mass and other space environment limitations not present for many terrestrial nuclear facilities. Potential accident conditions could expose nuclear power sources to extreme physical conditions. The Safety Framework is structured to provide guidance for both the programmatic and technical aspects of safety. In addition to sections containing specific guidance for governments and for management, it contains technical guidance pertinent to the design, development and all mission phases of space NPS applications. All sections of the Safety Framework contain elements directly relevant to engineers and space mission designers for missions involving space nuclear power sources. The challenge for organisations and engineers involved in the design and development processes of space nuclear power sources and applications is to implement the guidance provided in the Safety Framework by integrating it into the existing standard space mission infrastructure of design, development and operational requirements, practices and processes. This adds complexity to the standard space mission and launch approval processes. The Safety Framework is deliberately

  20. Impact of Fuel Failure on Criticality Safety of Used Nuclear Fuel

    SciTech Connect

    Marshall, William BJ J; Wagner, John C

    2012-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for considerably longer periods than originally intended (e.g., <40 years). Extended storage (ES) time and irradiation of nuclear fuel to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, can result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. This effort is primarily motivated by concerns related to the potential for fuel degradation during ES periods and transportation following ES. The criticality analyses consider representative UNF designs and cask systems and a range of fuel enrichments, burnups, and cooling times. The various failed-fuel configurations considered are designed to bound the anticipated effects of individual rod and general cladding failure, fuel rod deformation, loss of neutron absorber materials, degradation of canister internals, and gross assembly failure. The results quantify the potential impact on criticality safety associated with fuel reconfiguration and may be used to guide future research, design, and regulatory activities. Although it can be concluded that the criticality safety impacts of fuel reconfiguration during transportation subsequent to ES are manageable, the results indicate that certain configurations can result in a large increase in the effective neutron multiplication factor, k{sub eff}. Future work to inform decision making relative to which configurations are credible, and therefore need to be considered in a safety evaluation, is recommended.

  1. Nuclear Criticality Safety Calculational Analysis for Fissile Mass Limits and Spacing Requirements for 55 - Gallon Waste Drums

    SciTech Connect

    Davis, Thomas C.; Hesse, David J.; Tayloe, Jr., Robert W.

    1994-05-01

    A nuclear criticality safety analysis was performed to determine the fissile mass limits and spacing requirements for the storage of 55-gallon waste drums at the Portsmouth Gaseous Diffusion Plant (PORTS).

  2. Looking for residues involved in the muscle acylphosphatase catalytic mechanism and structural stabilization: role of Asn41, Thr42, and Thr46.

    PubMed

    Taddei, N; Stefani, M; Magherini, F; Chiti, F; Modesti, A; Raugei, G; Ramponi, G

    1996-06-04

    Asn41, Thr42, and Thr46 are invariant residues in both muscle and erythrocyte acylphosphatases isolated so far. Horse muscle acylphosphatase solution structure suggests their close spatial relationship to Arg23, the main substrate binding site. The catalytic and structural role of such residues, as well as their influence on muscle acylphosphatase stability, was investigated by preparing several gene mutants (Thr42Ala, Thr46Ala, Asn41Ala, Asn41Ser, and Asn41Gln) by oligonucleotide-directed mutagenesis. The mutated genes were cloned and expressed in Escherichia coli, and the mutant enzymes were purified by affinity chromatography and investigated as compared to the wild-type enzyme. The specific activity and substrate affinity of Thr42 and Thr46 mutants were not significantly affected. On the contrary, Asn41 mutants showed a residual negligible activity (about 0.05-0.15% as compared to wild-type enzyme), though maintaining an unchanged binding capability of both substrate and inorganic phosphate, an enzyme competitive inhibitor. According to the 1H nuclear magnetic resonance spectroscopy and circular dichroism results, all mutants elicited well-constrained native-like secondary and tertiary structures. Thermodynamic parameters, as calculated from circular dichroism data, demonstrated a significantly decreased stability of the Thr42 mutant under increasing temperatures and urea concentrations. The reported results strongly support a direct participation of Asn41 to the enzyme catalytic mechanism, indicating that Asn41 mutants may well represent a useful tool for the investigation of the enzyme physiological function by the negative dominant approach.

  3. Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Reactors and Safety Code Validation

    SciTech Connect

    Szabados, Laszlo

    2004-01-15

    The Paks nuclear power plant is equipped with pressurized water reactors of the VVER-440/213 type. These plants have a number of special features, namely, six-loop primary circuit, horizontal steam generators, loop seal in both hot and cold legs, safety injection tank setpoint pressure higher than secondary pressure, etc. As a consequence of the special design solutions, the transient behavior of such a reactor system is different from the usual pressurized water reactor system behavior. To study the transient behavior of these plants, the PMK-2 integral-type facility, a thermal-hydraulic model of the Paks nuclear power plant, was designed and constructed.A short description of the specific design solutions of the VVER-440/213-type plants is given with the modeling aspects and similarity criteria applied to the design of the PMK-2 facility. Since the startup of the facility in 1985, 48 experiments have been performed primarily in an international framework with the participation of several experts from European and overseas countries to study one- and two-phase natural circulation, loss-of-coolant accidents, special plant transients, and experiments in support of the accident management measures. The results of several experiments illustrate the system effects of special design solutions and the effectiveness of bleed-and-feed accident management measures. A brief commentary on the thermal-hydraulic system code validation is provided, and conclusions are offered.

  4. Hazards and hazard combinations relevant for the safety of nuclear power plants

    NASA Astrophysics Data System (ADS)

    Decker, Kurt; Brinkman, Hans; Raimond, Emmanuel

    2017-04-01

    The potential of the contemporaneous impact of different, yet causally related, hazardous events and event cascades on nuclear power plants is a major contributor to the overall risk of nuclear installations. In the aftermath of the Fukushima accident, which was caused by a combination of severe ground shaking by an earthquake, an earthquake-triggered tsunami and the disruption of the plants from the electrical grid by a seismically induced landslide, hazard combinations and hazard cascades moved into the focus of nuclear safety research. We therefore developed an exhaustive list of external hazards and hazard combinations which pose potential threats to nuclear installations in the framework of the European project ASAMPSAE (Advanced Safety Assessment: Extended PSA). The project gathers 31 partners from Europe, North Amerika and Japan. The list comprises of exhaustive lists of natural hazards, external man-made hazards, and a cross-correlation matrix of these hazards. The hazard list is regarded comprehensive by including all types of hazards that were previously cited in documents by IAEA, the Western European Nuclear Regulators Association (WENRA), and others. 73 natural hazards and 24 man-made external hazards are included. Natural hazards are grouped into seismotectonic hazards, flooding and hydrological hazards, extreme values of meteorological phenomena, rare meteorological phenomena, biological hazards / infestation, geological hazards, and forest fire / wild fire. The list of external man-made hazards includes industry accidents, military accidents, transportation accidents, pipeline accidents and other man-made external events. The large number of different hazards results in the extremely large number of 5.151 theoretically possible hazard combinations (not considering hazard cascades). In principle all of these combinations are possible to occur by random coincidence except for 82 hazard combinations that - depending on the time scale - are mutually

  5. 78 FR 68102 - Atomic Safety and Licensing Board; In the Matter of Nuclear Innovation North America LLC (South...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Atomic Safety and Licensing Board; In the Matter of Nuclear Innovation North America LLC (South Texas Project Units 3 and 4); Notice of Hearing (Application for Combined Licenses) November 6, 2013. Before Administrative Judges: Michael M....

  6. 76 FR 37799 - DOE Final Decision in Response to Recommendation 2010-1 of the Defense Nuclear Facilities Safety...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-28

    ... Indiana Avenue NW., Suite 700, Washington, DC 20004. FOR FURTHER INFORMATION CONTACT: Ms. Amanda Anderson... analysis. Attachment 1 February 28, 2011 The Honorable Peter S. Winokur, Chairman, Defense Nuclear... 2 May 27, 2011 The Honorable Peter S. Winokur, Chairman, Defense Nuclear Facilities Safety...

  7. Implementing Stakeholders' Access to Expertise: Experimenting on Nuclear Installations' Safety Cases - 12160

    SciTech Connect

    Gilli, Ludivine; Charron, Sylvie

    2012-07-01

    In 2009 and 2010, the Institute for Nuclear Safety and Radiation Protection (IRSN) led two pilot actions dealing with nuclear installations' safety cases. One concerned the periodical review of the French 900 MWe nuclear reactors, the other concerned the decommissioning of a workshop located on the site of Areva's La Hague fuel-reprocessing plant site in Northwestern France. The purpose of both these programs was to test ways for IRSN and a small number of stakeholders (Non-Governmental Organizations (NGOs) members, local elected officials, etc.) to engage in technical discussions. The discussions were intended to enable the stakeholders to review future applications and provide valuable input. The test cases confirmed there is a definite challenge in successfully opening a meaningful dialogue to discuss technical issues, in particular the fact that most expertise reports were not public and the conflict that exists between the contrary demands of transparency and confidentiality of information. The test case also confirmed there are ways to further improvement of stakeholders' involvement. (authors)

  8. Technical support for the Ukrainian State Committee for Nuclear Radiation Safety on specific waste issues

    SciTech Connect

    Little, C.A.

    1995-07-01

    The government of Ukraine, a now-independent former member of the Soviet Union, has asked the United States to assist its State Committee for Nuclear and Radiation Safety (SCNRS) in improving its regulatory control in technical fields for which it has responsibility. The US Nuclear Regulatory Commission (NRC) is providing this assistance in several areas, including management of radioactive waste and spent fuel. Radioactive wastes resulting from nuclear power plant operation, maintenance, and decommissioning must be stored and ultimately disposed of appropriately. In addition, radioactive residue from radioisotopes used in various industrial and medical applications must be managed. The objective of this program is to provide the Ukrainian SCNRS with the information it needs to establish regulatory control over uranium mining and milling activities in the Zheltye Vody (Yellow Waters) area and radioactive waste disposal in the Pripyat (Chernobyl) area among others. The author of this report, head of the Environmental Technology Section, Health Sciences Research Division of Oak Ridge National Laboratory, accompanied NRC staff to Ukraine to meet with SCNRS staff and visit sites in question. The report highlights problems at the sites visited and recommends license conditions that SCNRS can require to enhance safety of handling mining and milling wastes. The author`s responsibility was specifically for the visit to Zheltye Vody and the mining and milling waste sites associated with that facility. An itinerary for the Zheltye Vody portion of the trip is included as Appendix A.

  9. DOE, 2013. A Report to the Secretary of Energy: Beyond Design Basis Event Pilot Evaluations, Results and Recommendations for Improvements to Enhance Nuclear Safety at DOE Nuclear Facilities

    SciTech Connect

    None, None

    2013-02-01

    In the six months after the March 2011 Fukushima Daiichi nuclear power plant accident in Japan, the U.S. Department of Energy (DOE) took several actions to review the safety of its nuclear facilities and identify situations where near-term improvements could be made. These actions and recommendations were addressed in an August 2011 report to the Secretary of Energy, Review of Requirements and Capabilities for Analyzing and Responding to Beyond Design Basis Events. Based on recommendations in the August 2011 report, DOE embarked on a project to develop and refine guidance that supports improvements in DOE’s processes for analyzing and mitigating beyond design basis events (BDBEs), i.e., events such as earthquakes that are more severe than the events that formed the basis of the design for DOE’s nuclear facilities. The results of this BDBE project and recommendations for further DOE actions are provided in this follow-on report. The main activity of the BDBE project was the pilot application of guidance for evaluation of BDBE analysis and mitigation features at four DOE nuclear facilities representing a range of DOE sites, nuclear facility types/activities, and responsible program offices. The pilot evaluations focused on: (1) BDBE evaluations as documented in the facility Documented Safety Analysis (DSA); (2) potential BDBE vulnerabilities and margins to failure for facility safety features as obtained from general area and specific system walkdowns and design document reviews; and, (3) preparations made in facility and site emergency management programs to respond to severe accidents. The BDBE project also evaluated whether draft BDBE guidance on safety analysis and emergency management could be used to improve the analysis of, and preparations for, mitigating severe accidents and BDBEs. The pilot evaluation team paid close attention to related actions being pursued by the U.S. Nuclear Regulatory Commission, the U.S. commercial nuclear industry, and the

  10. Global Survey of the Concepts and Understanding of the Interfaces Between Nuclear Safety, Security, and Safeguards

    SciTech Connect

    Kovacic, Don N.; Stewart, Scott; Erickson, Alexa R.; Ford, Kerrie D.; Mladineo, Stephen V.

    2015-07-15

    There is increasing global discourse on how the elements of nuclear safety, security, and safeguards can be most effectively implemented in nuclear power programs. While each element is separate and unique, they must nevertheless all be addressed in a country’s laws and implemented via regulations and in facility operations. This topic is of particular interest to countries that are currently developing the infrastructure to support nuclear power programs. These countries want to better understand what is required by these elements and how they can manage the interfaces between them and take advantages of any synergies that may exist. They need practical examples and guidance in this area in order to develop better organizational strategies and technical capacities. This could simplify their legal, regulatory, and management structures and avoid inefficient approaches and costly mistakes that may not be apparent to them at this early stage of development. From the perspective of IAEA International Safeguards, supporting Member States in exploring such interfaces and synergies provides a benefit to them because it acknowledges that domestic safeguards in a country do not exist in a vacuum. Instead, it relies on a strong State System of Accounting and Control that is in turn dependent on a capable and independent regulatory body as well as a competent operator and technical staff. These organizations must account for and control nuclear material, communicate effectively, and manage and transmit complete and correct information to the IAEA in a timely manner. This, while in most cases also being responsible for the safety and security of their facilities. Seeking efficiencies in this process benefits international safeguards and nonproliferation. This paper will present the results of a global survey of current and anticipated approaches and practices by countries and organizations with current or future nuclear power programs on how they are implementing, or

  11. Validation of Nuclear Criticality Safety Software and 27 energy group ENDF/B-IV cross sections

    SciTech Connect

    Lee, B.L. Jr.

    1994-08-01

    The validation documented in this report is based on calculations that were executed during June through August 1992, and was completed in June 1993. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Martin Marietta Nuclear Criticality Safety Software. The first portion of the document outlines basic information in regard to validation of NCSS using ENDF/B-IV 27-group cross sections on the IBM 3090 at ORNL. A basic discussion of the NCSS system is provided, some discussion on the validation database and validation in general. Then follows a detailed description of the statistical analysis which was applied. The results of this validation indicate that the NCSS software may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant. When the validation results are treated as a single group, there is 95% confidence that 99.9% of future calculations of similar critical systems will have a calculated K{sub eff} > 0.9616. Based on this result the Portsmouth Nuclear Criticality Safety Department has adopted the calculational acceptance criteria that a k{sub eff} + 2{sigma} {le} 0.95 is safety subcritical. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25.

  12. Technical basis for environmental qualification of computer-based safety systems in nuclear power plants

    SciTech Connect

    Korsah, K.; Wood, R.T.; Tanaka, T.J.; Antonescu, C.E.

    1997-10-01

    This paper summarizes the results of research sponsored by the US Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. This research was conducted by the Oak Ridge National Laboratory (ORNL) and Sandia National Laboratories (SNL). ORNL investigated potential failure modes and vulnerabilities of microprocessor-based technologies to environmental stressors, including electromagnetic/radio-frequency interference, temperature, humidity, and smoke exposure. An experimental digital safety channel (EDSC) was constructed for the tests. SNL performed smoke exposure tests on digital components and circuit boards to determine failure mechanisms and the effect of different packaging techniques on smoke susceptibility. These studies are expected to provide recommendations for environmental qualification of digital safety systems by addressing the following: (1) adequacy of the present preferred test methods for qualification of digital I and C systems; (2) preferred standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging in qualification testing for equipment that is to be located in mild environments; and (5) determination of an appropriate approach to address smoke in a qualification program.

  13. Probabilistic Safety Assessment of External Flooding Protection for Nuclear Power Plants in Germany

    NASA Astrophysics Data System (ADS)

    Berg, Heinz Peter; Goertz, Rudolf; Froehmel, Thomas; Winter, Christian

    Methods to systematically analyse existing nuclear power plants (NPP) regarding the adequacy of their existing protection equipment against external hazards, e.g. flooding, can be of deterministic as well as probabilistic nature. In the past the adequacy of the protection measures has been assessed only on a deterministic basis. The German regulatory body has issued probabilistic safety assessment (PSA) guidelines, which had been elaborated for a comprehensive integrated safety review of all NPP in operation. Amongst others the guidelines imply, that probabilistic considerations regarding external flooding are required. This paper presents a newly developed graded approach for the probabilistic assessment of external flooding. Main aspects are explained such as the underlying probabilistic considerations and the mathematical procedures for the calculation of exceedance frequencies, which have recently been developed and issued as part of the German Nuclear Safety Standard. Exemplarily it has been investigated if extreme events such as tsunami waves could be a hazard for NPP at coastal sites in Germany. Here it could be shown that due to limited source mechanisms and the specific morphological conditions in the North Sea no dedicated measures for protection against tsunamis in the German Bight are necessary.

  14. WTEC monograph on instrumentation, control and safety systems of Canadian nuclear facilities

    NASA Technical Reports Server (NTRS)

    Uhrig, Robert E.; Carter, Richard J.

    1993-01-01

    This report updates a 1989-90 survey of advanced instrumentation and controls (I&C) technologies and associated human factors issues in the U.S. and Canadian nuclear industries carried out by a team from Oak Ridge National Laboratory (Carter and Uhrig 1990). The authors found that the most advanced I&C systems are in the Canadian CANDU plants, where the newest plant (Darlington) has digital systems in almost 100 percent of its control systems and in over 70 percent of its plant protection system. Increased emphasis on human factors and cognitive science in modern control rooms has resulted in a reduced workload for the operators and the elimination of many human errors. Automation implemented through digital instrumentation and control is effectively changing the role of the operator to that of a systems manager. The hypothesis that properly introducing digital systems increases safety is supported by the Canadian experience. The performance of these digital systems has been achieved using appropriate quality assurance programs for both hardware and software development. Recent regulatory authority review of the development of safety-critical software has resulted in the creation of isolated software modules with well defined interfaces and more formal structure in the software generation. The ability of digital systems to detect impending failures and initiate a fail-safe action is a significant safety issue that should be of special interest to nuclear utilities and regulatory authorities around the world.

  15. 76 FR 37798 - DOE Response to Recommendation 2010-2 of the Defense Nuclear Facilities Safety Board, Pulse Jet...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-28

    ... Response to Recommendation 2010-2 of the Defense Nuclear Facilities Safety Board, Pulse Jet Mixing at the..., concerning Pulse Jet Mixing at the Waste Treatment and Immobilization Plant, to the Department of Energy. In... Safety Board (Board) Recommendation 2010-2, Pulse Jet Mixing (PJM) at the Waste Treatment and...

  16. 78 FR 9902 - DOE Response to Recommendation 2012-2 of the Defense Nuclear Facilities Safety Board, Hanford...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-12

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF ENERGY DOE Response to Recommendation 2012-2 of the Defense Nuclear Facilities Safety Board, Hanford Tank Farms..., Hanford Tank Farms Flammable Gas Safety Strategy. This document corrects an error in that notice....

  17. 78 FR 47805 - Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-06

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing a revised regulatory guide (RG), Revision 1 of RG 1.170, ``Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants.'' This RG endorses the Institute of Electrical and Electronic Engineers (IEEE) Standard 829-2008, ``IEEE Standard for Software and System Test Documentation'' with the clarifications and exceptions stated in Section C, ``Staff Regulatory Guidance'' of the RG. IEEE Std. 829-2008 describes methods that the NRC considers acceptable for use in complying with NRC regulations for developing test documentation and design quality in the software used in safety systems of nuclear power plants.

  18. Nuclear safety considerations in the conceptual design of a fast reactor for space electric power and propulsion

    NASA Technical Reports Server (NTRS)

    Hsieh, T.-M.; Koenig, D. R.

    1977-01-01

    Some nuclear safety aspects of a 3.2 mWt heat pipe cooled fast reactor with out-of-core thermionic converters are discussed. Safety related characteristics of the design including a thin layer of B4C surrounding the core, the use of heat pipes and BeO reflector assembly, the elimination of fuel element bowing, etc., are highlighted. Potential supercriticality hazards and countermeasures are considered. Impacts of some safety guidelines of space transportation system are also briefly discussed, since the currently developing space shuttle would be used as the primary launch vehicle for the nuclear electric propulsion spacecraft.

  19. Transportable nuclear power plant TEC-M with two reactor plants of improved safety

    SciTech Connect

    Ogloblin, B.G.; Sazonov, A.G.; Svishchev, A.M.; Gromov, B.F.; Zelensky, V.N.; Komkova, O.I.; Sidorov, V.I.; Tolstopyatov, V.P.; Toshinsky, G.I.

    1993-12-31

    Liquid metals are the best to meet the requirements of inherently safety nuclear power plants among the coolants used. A great experience has been gained in lead coolant power plant development and operation as applied to transportable power set-ups. Low chemical activity of this coolant with respect to air-water interaction is a determining factor for this coolant. The transportable nuclear power plant is described. It is intended to generate electric power for populated areas placed a long distance from the main electric power supply sources where it is difficult or not economical to deliver the conventional types of fuel. There are several remote areas in Siberia, Kamchatka in need of this type of power plant.

  20. Use of artificial intelligence to enhance the safety of nuclear power plants

    SciTech Connect

    Uhrig, R.E.

    1988-01-01

    In the operation of a nuclear power plant, the sheer magnitude of the number of process parameters and systems interactions poses difficulties for the operators, particularly during abnormal or emergency situations. Recovery from an upset situation depends upon the facility with which the available raw data can be converted into and assimilated as meaningful knowledge. Plant personnel are sometimes affected by stress and emotion, which may have varying degrees of influence on their performance. Expert systems can take some of the uncertainty and guesswork out of their decisions by providing expert advice and rapid access to a large information base. Application of artificial intelligence technologies, particularly expert systems, to control room activities in a nuclear power plant has the potential to reduce operator error and improve power plant safety and reliability. 12 refs.

  1. Operation Grenadier. Onsite radiological safety report for announced nuclear tests, October 1984-September 1985

    SciTech Connect

    Mullen, O.W.; Eubank, B.F.

    1986-09-01

    Grenadier was the name assigned to the series of underground nuclear experiments conducted at the Nevada Test Site from October 1, 1984 through September 30, 1985. This report includes those experiments publicly announced. Remote radiation measurements were taken during and after each nuclear experiment by a telemetry system. Monitors with portable radiation detection instruments surveyed reentry routes into ground zeros before other planned entries were made. Continuous surveillance was provided while personnel were in radiation areas and appropriate precautions were taken to protect persons from unnecessary exposure to radiation and toxic gases. Protective clothing and equipment were issued as needed. Complete radiological safety and industrial hygiene coverage was provided during drilling and mineback operations. Telemetered and portable radiation detector measurements are listed. Detection instrumentation used is described and specific operational procedures are defined.

  2. Operation Cornerstone onsite radiological safety report for announced nuclear tests, October 1988--September 1989

    SciTech Connect

    Not Available

    1990-08-01

    Cornerstone was the name assigned to the series of underground nuclear experiments conducted at the Nevada Test Site (NTS) from October 1, 1988, through September 30, 1989. This report includes those experiments publicly announced. Remote radiation measurements were taken during and after each nuclear experiment by a telemetry system. Radiation Protection Technicians (RPT) with portable radiation detection instruments surveyed reentry routes into ground zeros (GZ) before other planned entries were made. Continuous surveillance was provided while personnel were in radiation areas and appropriate precautions were taken to protect persons from unnecessary exposure to radiation and toxic gases. Protective clothing and equipment were issued as needed. Complete radiological safety and industrial hygiene coverage were provided during drilling and mineback operations. Telemetered and portable radiation detector measurements are listed. Detection instrumentation used is described and specific operational procedures are defined.

  3. Optical fiber sensors to improve the safety of nuclear power plants

    NASA Astrophysics Data System (ADS)

    Ferdinand, P.; Magne, S.; Laffont, G.

    2013-09-01

    Safety must always prevail in Nuclear Power Plants (NPPs), as shown at Fukushima-Daiichi. So, innovations are clearly needed to strengthen instrumentations, which went inoperative during this nuclear accident as a consequence of power supply losses. Possible improvements concern materials and structures, which may be remotely monitored thanks to Optical Fiber Sensors (OFS). We detail topics involving OFS helpful for monitoring, in nominal conditions as well as during a severe accident. They include distributed sensing (Rayleigh, Raman, Brillouin) for both temperature sensing and structure monitoring as well as H2 concentration and ionizing radiation monitoring. For future plants, Fiber Bragg Grating (FBG) sensors are considered up to high temperature for sodium-cooled fast reactor monitoring. These applications can benefit from fiber advantages: sensor multiplexing, multi-km range, no risk-to-people, no common failure mode with other technologies, remote sensing, and the ability to operate in case of power supply lost in the NPP.

  4. Safety of interim storage solutions of used nuclear fuel during extended term

    SciTech Connect

    Shelton, C.; Bader, S.; Issard, H.; Arslan, M.

    2013-07-01

    In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

  5. Guidelines for nuclear power plant safety issue prioritization information development. Supplement 3

    SciTech Connect

    Andrews, W.B.; Bickford, W.E.; Counts, C.A.; Gallucci, R.H.V.; Heaberlin, S.W.; Powers, T.B.; Weakley, S.A.

    1985-09-01

    This supplemental report is the fourth in a series that document and use methods developed to calculate, for prioritization purposes, the risk, dose and cost impacts of implementing resolutions to reactor safety issues. The initial report in this series was published by Andrews et al. in 1983 as NUREG/CR-2800. This supplement consists of two parts describing separate research efforts: (1) an alternative human factors methodology approach, and (2) a prioritization of the NRC's Human Factors Program Plan. The alternative human factors methodology approach may be used in specific future cases in which the methods identified in the initial report (NUREG/CR-2800) may not adequately assess the proper impact for resolution of new safety issues. The alternative methodology included in this supplement is entitled ''Methodology for Estimating the Public Risk Reduction Affected by Human Factors Improvement.'' The prioritization section of this report is entitled ''Prioritization of the US Nuclear Regulatory Commission Human Factors Program Plan.''

  6. Institutional implications of establishing safety goals for nuclear power plants. [PWR; BWR

    SciTech Connect

    Morris, F.A.; Hooper, R.L.

    1983-07-01

    The purpose of this project is to anticipate and address institutional problems that may arise from the adoption of NRC's proposed Policy Statement on Safety Goals for Nuclear Power Plants. The report emphasizes one particular category of institutional problems: the possible use of safety goals as a basis for legal challenges to NRC actions, and the resolution of such challenges by the courts. Three types of legal issues are identified and analyzed. These are, first, general legal issues such as access to the legal system, burden of proof, and standard of proof. Second is the particular formulation of goals. Involved here are such questions as sustainable rationale, definitions, avoided issues, vagueness of time and space details, and degree of conservatism. Implementation brings up the third set of issues which include interpretation and application, linkage to probabilistic risk assessment, consequences as compared to events, and the use of results.

  7. Techniques to evaluate the importance of common cause degradation on reliability and safety of nuclear weapons.

    SciTech Connect

    Darby, John L.

    2011-05-01

    As the nuclear weapon stockpile ages, there is increased concern about common degradation ultimately leading to common cause failure of multiple weapons that could significantly impact reliability or safety. Current acceptable limits for the reliability and safety of a weapon are based on upper limits on the probability of failure of an individual item, assuming that failures among items are independent. We expanded the current acceptable limits to apply to situations with common cause failure. Then, we developed a simple screening process to quickly assess the importance of observed common degradation for both reliability and safety to determine if further action is necessary. The screening process conservatively assumes that common degradation is common cause failure. For a population with between 100 and 5000 items we applied the screening process and conclude the following. In general, for a reliability requirement specified in the Military Characteristics (MCs) for a specific weapon system, common degradation is of concern if more than 100(1-x)% of the weapons are susceptible to common degradation, where x is the required reliability expressed as a fraction. Common degradation is of concern for the safety of a weapon subsystem if more than 0.1% of the population is susceptible to common degradation. Common degradation is of concern for the safety of a weapon component or overall weapon system if two or more components/weapons in the population are susceptible to degradation. Finally, we developed a technique for detailed evaluation of common degradation leading to common cause failure for situations that are determined to be of concern using the screening process. The detailed evaluation requires that best estimates of common cause and independent failure probabilities be produced. Using these techniques, observed common degradation can be evaluated for effects on reliability and safety.

  8. Conduct and results of the Interagency Nuclear Safety Review Panel's evaluation of the Ulysses space mission

    NASA Technical Reports Server (NTRS)

    Sholtis, Joseph A., Jr.; Huff, Darrell A.; Gray, Leven B.; Klug, Norman P.; Winchester, Robert O.

    1991-01-01

    After over a year of detailed independent review, a Safety Evaluation Report (SER) was prepared for the nuclear-powered Ulysses spacecraft mission. An overview is presently given of the Ulysses mission as well as the conduct and results of the evaluation process; in the final analysis, the SER was supportive of a positive launch decision. Eleven key accident scenarios were carried through to complete analysis, out of a total of 19 possible accidents involving a potential for fuel release to the spacecraft environment.

  9. Determining a cost/effectiveness/safety tradeoff methodology for strategic nuclear warheads

    SciTech Connect

    Erickson, S.A. Jr.; Hall, C.H.

    1992-04-27

    Department of Energy national laboratories are charged with anticipating with a long leadtime which technologies for nuclear warheads should be developed. The Safe Warhead System Study was constituted to provide Lawrence Livermore National Laboratory management with information and suggestions for making such decisions for enhanced safety warheads. The Minuteman III replacement warheads were analyzed as a test case and that information was used to identify and describe the dominant issues, to develop a methodology and to make initial recommendations. The test case work resulted in several insights into how ongoing design and engineering interacts with the technology ranking and on how to cope with the ubiquitous uncertainties relating to our current ICBM force.

  10. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    SciTech Connect

    Koponen, B.L.; Hampel, V.E.

    1982-10-21

    This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains-in chronological order-the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41.

  11. Style, content and format guide for writing safety analysis documents: Volume 2, Safety assessment reports for DOE non-nuclear facilities

    SciTech Connect

    Mahn, J.A.; Silver, R.C.; Balas, Y.; Gilmore, W.

    1995-07-01

    The purpose of Volume 2 of this 4-volume style guide is to furnish guidelines on writing and publishing Safety Assessment Reports (SAs) for DOE non-nuclear facilities at Sandia National Laboratories. The scope of Volume 2 encompasses not only the general guidelines for writing and publishing, but also the prescribed topics/appendices contents along with examples from typical SAs for DOE non-nuclear facilities.

  12. Probabilistic cost-benefit analysis of enhanced safety features for strategic nuclear weapons at a representative location

    SciTech Connect

    Stephens, D.R.; Hall, C.H.; Holman, G.S.; Graham, K.F.; Harvey, T.F.; Serduke, F.J.D.

    1993-10-01

    We carried out a demonstration analysis of the value of developing and implementing enhanced safety features for nuclear weapons in the US stockpile. We modified an approach that the Nuclear Regulatory Commission (NRC) developed in response to a congressional directive that NRC assess the ``value-impact`` of regulatory actions for commercial nuclear power plants. Because improving weapon safety shares some basic objectives with NRC regulations, i.e., protecting public health and safety from the effects of accidents involving radioactive materials, we believe the NRC approach to be appropriate for evaluating weapons-safety cost-benefit issues. Impact analysis includes not only direct costs associated with retrofitting the weapon system, but also the expected costs (or economic risks) that are avoided by the action, i.e., the benefits.

  13. Annual report to Congress. Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, calendar year 2000

    SciTech Connect

    2001-03-01

    This Annual Report to the Congress describes the Department of Energy's activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board. During 2000, the Department completed its implementation and proposed closure of one Board recommendation and completed all implementation plan milestones associated with two additional Board recommendations. Also in 2000, the Department formally accepted two new Board recommendations and developed implementation plans in response to those recommendations. The Department also made significant progress with a number of broad-based safety initiatives. These include initial implementation of integrated safety management at field sites and within headquarters program offices, issuance of a nuclear safety rule, and continued progress on stabilizing excess nuclear materials to achieve significant risk reduction.

  14. Integration of the advanced transparency framework to advanced nuclear systems : enhancing Safety, Operations, Security and Safeguards (SOSS).

    SciTech Connect

    Mendez, Carmen Margarita; Rochau, Gary Eugene; Cleary, Virginia D.

    2008-08-01

    The advent of the nuclear renaissance gives rise to a concern for the effective design of nuclear fuel cycle systems that are safe, secure, nonproliferating and cost-effective. We propose to integrate the monitoring of the four major factors of nuclear facilities by focusing on the interactions between Safeguards, Operations, Security, and Safety (SOSS). We proposed to develop a framework that monitors process information continuously and can demonstrate the ability to enhance safety, operations, security, and safeguards by measuring and reducing relevant SOSS risks, thus ensuring the safe and legitimate use of the nuclear fuel cycle facility. A real-time comparison between expected and observed operations provides the foundation for the calculation of SOSS risk. The automation of new nuclear facilities requiring minimal manual operation provides an opportunity to utilize the abundance of process information for monitoring SOSS risk. A framework that monitors process information continuously can lead to greater transparency of nuclear fuel cycle activities and can demonstrate the ability to enhance the safety, operations, security and safeguards associated with the functioning of the nuclear fuel cycle facility. Sandia National Laboratories (SNL) has developed a risk algorithm for safeguards and is in the process of demonstrating the ability to monitor operational signals in real-time though a cooperative research project with the Japan Atomic Energy Agency (JAEA). The risk algorithms for safety, operations and security are under development. The next stage of this work will be to integrate the four algorithms into a single framework.

  15. Application of Framework for Integrating Safety, Security and Safeguards (3Ss) into the Design Of Used Nuclear Fuel Storage Facility

    SciTech Connect

    Badwan, Faris M.; Demuth, Scott F

    2015-01-06

    Department of Energy’s Office of Nuclear Energy, Fuel Cycle Research and Development develops options to the current commercial fuel cycle management strategy to enable the safe, secure, economic, and sustainable expansion of nuclear energy while minimizing proliferation risks by conducting research and development focused on used nuclear fuel recycling and waste management to meet U.S. needs. Used nuclear fuel is currently stored onsite in either wet pools or in dry storage systems, with disposal envisioned in interim storage facility and, ultimately, in a deep-mined geologic repository. The safe management and disposition of used nuclear fuel and/or nuclear waste is a fundamental aspect of any nuclear fuel cycle. Integrating safety, security, and safeguards (3Ss) fully in the early stages of the design process for a new nuclear facility has the potential to effectively minimize safety, proliferation, and security risks. The 3Ss integration framework could become the new national and international norm and the standard process for designing future nuclear facilities. The purpose of this report is to develop a framework for integrating the safety, security and safeguards concept into the design of Used Nuclear Fuel Storage Facility (UNFSF). The primary focus is on integration of safeguards and security into the UNFSF based on the existing Nuclear Regulatory Commission (NRC) approach to addressing the safety/security interface (10 CFR 73.58 and Regulatory Guide 5.73) for nuclear power plants. The methodology used for adaptation of the NRC safety/security interface will be used as the basis for development of the safeguards /security interface and later will be used as the basis for development of safety and safeguards interface. Then this will complete the integration cycle of safety, security, and safeguards. The overall methodology for integration of 3Ss will be proposed, but only the integration of safeguards and security will be applied to the design of the

  16. Pacific Northwest Laboratory: Annual report for 1986 to the Assistant Secretary for Environment, Safety and Health: Part 5, Nuclear and operational safety

    SciTech Connect

    Faust, L.G.; Kennedy, W.E.; Steelman, B.L.; Selby, J.M.

    1987-02-01

    Part 5 of the 1986 Annual Report to the Department of Energy's Assistant Secretary for Environment, Safety and Health presents Pacific Northwest Laboratory's progress on work performed for the Office of Nuclear Safety, the Office of Operational Safety, and for the Office of Environmental Analysis. For each project, as identified by the Field Task Proposal/Agreement, articles describe progress made during fiscal year 1986. Authors of these articles represent a broad spectrum of capabilities derived from three of the seven research departments of the Laboratory, reflecting the interdisciplinary nature of the work.

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

    SciTech Connect

    J. Blair Briggs; Anatoly Tsibulya; Yevgeniy Rozhikhin

    2012-03-01

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

  18. Evaluation of natural phenomena hazards as part of safety assessments for nuclear facilities

    SciTech Connect

    Kot, C.A.; Hsieh, B.J.; Srinivasan, M.G.; Shin, Y.W.

    1995-02-01

    The continued operation of existing US Department of Energy (DOE) nuclear facilities and laboratories requires a safety reassessment based on current criteria and guidelines. This also includes evaluations for the effects of Natural Phenomena Hazards (NPH), for which these facilities may not have been designed. The NPH evaluations follow the requirements of DOE Order 5480.28, Natural Phenomena Hazards Mitigation (1993) which establishes NPH Performance Categories (PCs) for DOE facilities and associated target probabilistic performance goals. These goals are expressed as the mean annual probability of exceedance of acceptable behavior for structures, systems and components (SSCs) subjected to NPH effects. The assignment of an NPH Performance Category is based on the overall hazard categorization (low, moderate, high) of a facility and on the function of an SSC under evaluation (DOE-STD-1021, 1992). Detailed guidance for the NPH analysis and evaluation criteria are also provided (DOE-STD-1020, 1994). These analyses can be very resource intensive, and may not be necessary for the evaluation of all SSCs in existing facilities, in particular for low hazard category facilities. An approach relying heavily on screening inspections, engineering judgment and use of NPH experience data (S. J. Eder et al., 1993), can minimize the analytical effort, give reasonable estimates of the NPH susceptibilities, and yield adequate information for an overall safety evaluation of the facility. In the following sections this approach is described in more detail and is illustrated by an application to a nuclear laboratory complex.

  19. Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants

    SciTech Connect

    O, J.M.; Higgins, J.; Stephen Fleger - NRC

    2011-09-19

    The U.S. Nuclear Regulatory Commission (NRC) reviews the human factors engineering (HFE) programs of applicants for nuclear power plant construction permits, operating licenses, standard design certifications, and combined operating licenses. The purpose of these safety reviews is to help ensure that personnel performance and reliability are appropriately supported. Detailed design review procedures and guidance for the evaluations is provided in three key documents: the Standard Review Plan (NUREG-0800), the HFE Program Review Model (NUREG-0711), and the Human-System Interface Design Review Guidelines (NUREG-0700). These documents were last revised in 2007, 2004 and 2002, respectively. The NRC is committed to the periodic update and improvement of the guidance to ensure that it remains a state-of-the-art design evaluation tool. To this end, the NRC is updating its guidance to stay current with recent research on human performance, advances in HFE methods and tools, and new technology being employed in plant and control room design. This paper describes the role of HFE guidelines in the safety review process and the content of the key HFE guidelines used. Then we will present the methodology used to develop HFE guidance and update these documents, and describe the current status of the update program.

  20. Exploring Operational Safeguards, Safety, and Security by Design to Address Real Time Threats in Nuclear Facilities

    SciTech Connect

    Schanfein, Mark J.; Mladineo, Stephen V.

    2015-07-07

    Over the last few years, significant attention has been paid to both encourage application and provide domestic and international guidance for designing in safeguards and security in new facilities.1,2,3 However, once a facility is operational, safeguards, security, and safety often operate as separate entities that support facility operations. This separation is potentially a serious weakness should insider or outsider threats become a reality.Situations may arise where safeguards detects a possible loss of material in a facility. Will they notify security so they can, for example, check perimeter doors for tampering? Not doing so might give the advantage to an insider who has already, or is about to, move nuclear material outside the facility building. If outsiders break into a facility, the availability of any information to coordinate the facility’s response through segregated alarm stations or a failure to include all available radiation sensors, such as safety’s criticality monitors can give the advantage to the adversary who might know to disable camera systems, but would most likely be unaware of other highly relevant sensors in a nuclear facility.This paper will briefly explore operational safeguards, safety, and security by design (3S) at a high level for domestic and State facilities, identify possible weaknesses, and propose future administrative and technical methods, to strengthen the facility system’s response to threats.

  1. Evaluating the Cost, Safety, and Proliferation Risks of Small Floating Nuclear Reactors.

    PubMed

    Ford, Michael J; Abdulla, Ahmed; Morgan, M Granger

    2017-01-17

    It is hard to see how our energy system can be decarbonized if the world abandons nuclear power, but equally hard to introduce the technology in nonnuclear energy states. This is especially true in countries with limited technical, institutional, and regulatory capabilities, where safety and proliferation concerns are acute. Given the need to achieve serious emissions mitigation by mid-century, and the multidecadal effort required to develop robust nuclear governance institutions, we must look to other models that might facilitate nuclear plant deployment while mitigating the technology's risks. One such deployment paradigm is the build-own-operate-return model. Because returning small land-based reactors containing spent fuel is infeasible, we evaluate the cost, safety, and proliferation risks of a system in which small modular reactors are manufactured in a factory, and then deployed to a customer nation on a floating platform. This floating small modular reactor would be owned and operated by a single entity and returned unopened to the developed state for refueling. We developed a decision model that allows for a comparison of floating and land-based alternatives considering key International Atomic Energy Agency plant-siting criteria. Abandoning onsite refueling is beneficial, and floating reactors built in a central facility can potentially reduce the risk of cost overruns and the consequences of accidents. However, if the floating platform must be built to military-grade specifications, then the cost would be much higher than a land-based system. The analysis tool presented is flexible, and can assist planners in determining the scope of risks and uncertainty associated with different deployment options. © 2017 Society for Risk Analysis.

  2. 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

  3. Development of a Method for Quantifying the Reliability of Nuclear Safety-Related Software

    SciTech Connect

    Yi Zhang; Michael W. Golay

    2003-10-01

    The work of our project is intended to help introducing digital technologies into nuclear power into nuclear power plant safety related software applications. In our project we utilize a combination of modern software engineering methods: design process discipline and feedback, formal methods, automated computer aided software engineering tools, automatic code generation, and extensive feasible structure flow path testing to improve software quality. The tactics include ensuring that the software structure is kept simple, permitting routine testing during design development, permitting extensive finished product testing in the input data space of most likely service and using test-based Bayesian updating to estimate the probability that a random software input will encounter an error upon execution. From the results obtained the software reliability can be both improved and its value estimated. Hopefully our success in the project's work can aid the transition of the nuclear enterprise into the modern information world. In our work, we have been using the proprietary sample software, the digital Signal Validation Algorithm (SVA), provided by Westinghouse. Also our work is being done with their collaboration. The SVA software is used for selecting the plant instrumentation signal set which is to be used as the input the digital Plant Protection System (PPS). This is the system that automatically decides whether to trip the reactor. In our work, we are using -001 computer assisted software engineering (CASE) tool of Hamilton Technologies Inc. This tool is capable of stating the syntactic structure of a program reflecting its state requirements, logical functions and data structure.

  4. Technical Support Section Instrument Support Program for nuclear and nonnuclear facilities with safety requirements

    SciTech Connect

    Adkisson, B.P.; Allison, K.L.

    1995-01-01

    This document describes requirements, procedures, and supervisory responsibilities of the Oak Ridge National Laboratory (ORNL) Instrumentation and Controls (I&C) Division`s Technical Support Section (TSS) for instrument surveillance and maintenance in nonreactor nuclear facilities having identified Operational Safety Requirements (OSRs) or Limiting Conditions Document (LCDs). Implementation of requirements comply with the requirements of U.S. Department of Energy (DOE) Orders 5480.5, 5480.22, and 5481.1B; Martin Marietta Energy Systems, Inc. (Energy Systems), Policy Procedure ESS-FS-201; and ORNL SPP X-ESH-15. OSRs and LCDs constitute an agreement or contract between DOE and the facility operating management regarding the safe operation of the facility. One basic difference between OSRs and LCDs is that violation of an OSR is considered a Category II occurrence, whereas violation of an LCD requirement is considered a Category III occurrence (see Energy Systems Standard ESS-OP-301 and ORNL SPP X-GP-13). OSRs are required for high- and moderate-hazard nuclear facilities, whereas the less-rigorous LCDs are required for low-hazard nuclear facilities and selected {open_quotes}generally accepted{close_quotes} operations. Hazard classifications are determined through a hazard screening process, which each division conducts for its facilities.

  5. Electric Power quality Analysis in research reactor: Impacts on nuclear safety assessment and electrical distribution reliability

    SciTech Connect

    Touati, Said; Chennai, Salim; Souli, Aissa

    2015-07-01

    The increased requirements on supervision, control, and performance in modern power systems make power quality monitoring a common practise for utilities. Large databases are created and automatic processing of the data is required for fast and effective use of the available information. Aim of the work presented in this paper is the development of tools for analysis of monitoring power quality data and in particular measurements of voltage and currents in various level of electrical power distribution. The study is extended to evaluate the reliability of the electrical system in nuclear plant. Power Quality is a measure of how well a system supports reliable operation of its loads. A power disturbance or event can involve voltage, current, or frequency. Power disturbances can originate in consumer power systems, consumer loads, or the utility. The effect of power quality problems is the loss power supply leading to severe damage to equipments. So, we try to track and improve system reliability. The assessment can be focused on the study of impact of short circuits on the system, harmonics distortion, power factor improvement and effects of transient disturbances on the Electrical System during motor starting and power system fault conditions. We focus also on the review of the Electrical System design against the Nuclear Directorate Safety Assessment principles, including those extended during the last Fukushima nuclear accident. The simplified configuration of the required system can be extended from this simple scheme. To achieve these studies, we have used a demo ETAP power station software for several simulations. (authors)

  6. Consequence modeling for nuclear weapons probabilistic cost/benefit analyses of safety retrofits

    SciTech Connect

    Harvey, T.F.; Peters, L.; Serduke, F.J.D.; Hall, C.; Stephens, D.R.

    1998-01-01

    The consequence models used in former studies of costs and benefits of enhanced safety retrofits are considered for (1) fuel fires; (2) non-nuclear detonations; and, (3) unintended nuclear detonations. Estimates of consequences were made using a representative accident location, i.e., an assumed mixed suburban-rural site. We have explicitly quantified land- use impacts and human-health effects (e.g. , prompt fatalities, prompt injuries, latent cancer fatalities, low- levels of radiation exposure, and clean-up areas). Uncertainty in the wind direction is quantified and used in a Monte Carlo calculation to estimate a range of results for a fuel fire with uncertain respirable amounts of released Pu. We define a nuclear source term and discuss damage levels of concern. Ranges of damages are estimated by quantifying health impacts and property damages. We discuss our dispersal and prompt effects models in some detail. The models used to loft the Pu and fission products and their particle sizes are emphasized.

  7. Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization

    SciTech Connect

    Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Michael Pernice; Robert Nourgaliev

    2013-05-01

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

  8. On the Seismic Safety of Nuclear Power Plant Sites in South Korea

    NASA Astrophysics Data System (ADS)

    Choi, H.; Park, S.; Yang, J.; Shim, T.; Im, C. B.

    2016-12-01

    The Korean Peninsula is located at the far eastern part of Eurasian Plate, and within the intra-plate region several hundred km away from the nearest plate boundary. The earthquakes around the Korean Peninsula show the typical characteristics of intra-plate earthquakes. So to speak, those are low seismicity, relatively smaller magnitude than that of inter-plate earthquakes, and spatially irregular epicenters. There are 24 nuclear power plants (NPPs) in operation, 4 NPPs in completion of construction, and 4 NPPs in preparation of construction in South Korea. Even though the seismicity of the Korean Peninsula is known as relatively low, but because there are more than 30 NPPs within not so large territory, thorough the preparedness of NPPs' safety against earthquakes is required. The earthquake preparedness of NPPs in South Korea is composed of 4 stages: site election, design, construction and operation. Since regulatory codes and standards are strictly applied in each stage, the NPPs in South Korea are believed to be safe enough against the maximum potential earthquake ground motion. Through data analysis on geological and seismological characteristics of the region within a radius of 320 km from the site and the detailed geological survey of the area within a radius of 8 km from the site, the design earthquake ground motion of NPPs in South Korea is determined to be 0.2g (in case of newly constructed NPPs is 0.3g) considering the maximum potential earthquake ground motion and some safety margin. The ground motions and surface deformation caused by capable faults are also considered in the seismic design of NPPs. In addition, the Korea Institute of Nuclear Safety as a regulatory technical expert organization, has been operating independent real time earthquake monitoring network as a part of securing the seismic safety of NPP sites in South Korea since late 1990's. If earthquakes with more than magnitude 3.0 are occurred in the Korean Peninsula or the peak ground

  9. Safety control of nuclear power operations using self-programming neural networks

    SciTech Connect

    Jouse, W.C.; Williams, J.G. )

    1993-05-01

    In the design and operation of nuclear reactors, safety-related goals must be embedded in complex multivariate control strategies. It is often the case that the goals exist only as mental models in the mind of the designer or the operator. In order to effect control that is risk averse, the goals must be translated into an effective control strategy that can be both verified and validated. The relation that these safety goals have to a particular architecture of artificial neural network, the Barto-Sutton architecture, is examined and the capability of the network to embed safety goals in nontrivial control tasks is demonstrated. To realize these goals, the network was extended to encompass a multiple-input/multiple-output control structure. The network synthesizes a control schedule through the construction of artificial precursors to failure; these serve as an additional, virtual layer in the defenses against fission product release. The synthesized schedule can be visually inspected for anomalies and inconsistencies and is validated during training.

  10. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    NASA Astrophysics Data System (ADS)

    Galvez, Cristhian

    2011-12-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the passive safety cooling system with a dual purpose, to assess the capacity to maintain the core at safe temperatures and to assist the design process of this system to achieve this objective. The analysis requires the use of complex computational tools for simulation and verification using analytical solutions and comparisons with experimental data. This investigation builds upon previous detailed design work for the PB-AHTR components, including the core, reactivity control mechanisms and the intermediate heat exchanger, developed in 2008. In addition the study of this reference plant design employs a wealth of auxiliary information including thermal-hydraulic physical phenomena correlations for multiple geometries and thermophysical properties for the constituents of the plant. Finally, the set of performance requirements and limitations imposed from physical constrains and safety considerations provide with a criteria and metrics for acceptability of the design. The passive safety cooling system concept is turned into a detailed design as a result from this study. A methodology for the design of air-cooled passive safety systems was developed and a transient analysis of the plant, evaluating a scrammed loss of forced cooling event was performed. Furthermore, a design optimization study of the passive safety system and an approach for the validation and verification of the analysis is presented. This study demonstrates that the resulting point design responds properly to the

  11. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect

    Banerjee, Kaushik; Scaglione, John M

    2015-01-01

    The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated keff margin varies from 0.05 to almost 0.3 Δkeff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  12. Characterisation of Liquefaction Effects for Beyond-Design Basis Safety Assessment of Nuclear Power Plants

    NASA Astrophysics Data System (ADS)

    Bán, Zoltán; Győri, Erzsébet; János Katona, Tamás; Tóth, László

    2015-04-01

    -tree procedure. Earlier studies have shown that the potentially liquefiable layer at Paks Nuclear Power Plant is situated in relatively large depth. Therefore the applicability and adequacy of the methods at high overburden pressure is important. In case of existing facilities, the geotechnical data gained before construction aren't sufficient for the comprehensive liquefaction analysis. Performance of new geotechnical survey is limited. Consequently, the availability of the data has to be accounted while selection the analysis methods. Considerations have to be made for dealing with aleatory uncertainty related to the knowledge of the soil conditions. It is shown in the paper, a careful comparison and analysis of the results obtained by different methodologies provides the basis of the selection of practicable methods for the safety analysis of nuclear power plant for beyond design basis liquefaction hazard.

  13. Safety research programs sponsored by Office of Nuclear Regulatory Research: Progress report, January 1--June 30, 1988

    SciTech Connect

    Baum, J W; Boccio, J L; Diamond, D; Fitzpatrick, R; Ginsberg, T; Greene, G A; Guppy, J G; Hall, R E; Higgins, J C; Weiss, A J

    1988-12-01

    This progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Regulatory Applications, Division of Engineering, Division of Safety Issue Resolution, and Division of Systems Research of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research following the reorganization in July 1988. The previous reports have covered the period October 1, 1976 through December 31, 1987.

  14. Safety research programs sponsored by Office of Nuclear Regulatory Research: Progress report, July 1--September 30, 1988

    SciTech Connect

    Weiss, A J

    1989-02-01

    This progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Regulatory Applications, Division of Engineering, Division of Safety Issue Resolution, and Division of Systems of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research following the reorganization in July 1988. The previous reports have covered the period October 1, 1976 through June 30, 1988. 71 figs., 24 tabs.

  15. Safety research programs sponsored by Office of Nuclear Regulatory Research: Progress report, January 1--March 31, 1989

    SciTech Connect

    Weiss, A.J.

    1989-08-01

    This progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Regulatory Applications, Division of Engineering, Division of Safety Issue Resolution, and Division of Systems Research of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research following the reorganization in July 1988. The previous reports have covered the period October 1, 1976 through December 31, 1988.

  16. Safety research programs sponsored by Office of Nuclear Regulatory Research: Progress report, October 1--December 31, 1988

    SciTech Connect

    Weiss, A J; Azarm, A; Baum, J W; Boccio, J L; Carew, J; Diamond, D J; Fitzpatrick, R; Ginsberg, T; Greene, G A; Guppy, J G; Haber, S B

    1989-07-01

    This progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Regulatory Applications, Division of Engineering, Division of Safety Issue Resolution, and Division of Systems Research of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research following the reorganization in July 1988. The previous reports have covered the period October 1, 1976 through September 30, 1988.

  17. Annual report to Congress: Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, calendar year 1998

    SciTech Connect

    1999-02-01

    This is the ninth Annual Report to the Congress describing Department of Energy (Department) activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board (Board). The Board, an independent executive-branch agency established in 1988, provides advice and recommendations to the Secretary of energy regarding public health and safety issues at the Department`s defense nuclear facilities. The Board also reviews and evaluates the content and implementation of health and safety standards, as well as other requirements, relating to the design, construction, operation, and decommissioning of the Department`s defense nuclear facilities. The locations of the major Department facilities are provided. During 1998, Departmental activities resulted in the proposed closure of one Board recommendation. In addition, the Department has completed all implementation plan milestones associated with four other Board recommendations. Two new Board recommendations were received and accepted by the Department in 1998, and two new implementation plans are being developed to address these recommendations. The Department has also made significant progress with a number of broad-based initiatives to improve safety. These include expanded implementation of integrated safety management at field sites, a renewed effort to increase the technical capabilities of the federal workforce, and a revised plan for stabilizing excess nuclear materials to achieve significant risk reduction.

  18. Contribution to the safety assessment of instrumentation and control software for nuclear power plants: Application to SPIN N4

    SciTech Connect

    Soubies, B.; Henry, J.Y.; Le Meur, M.

    1995-04-01

    1300 MWe pressurised water reactors (PWRs), like the 1400 MWe reactors, operate with microprocessor-based safety systems. This is particularly the case for the Digital Integrated Protection System (SPIN), which trips the reactor in an emergency and sets in action the safeguard functions. The softwares used in these systems must therefore be highly dependable in the execution of their functions. In the case of SPIN, three players are working at different levels to achieve this goal: the protection system manufacturer, Merlin Gerin; the designer of the nuclear steam supply system, Framatome; the operator of the nuclear power plants, Electricite de France (EDF), which is also responsible for the safety of its installations. Regulatory licenses are issued by the French safety authority, the Nuclear Installations Safety Directorate (French abbreviation DSIN), subsequent to a successful examination of the technical provisions adopted by the operator. This examination is carried out by the IPSN and the standing group on nuclear reactors. This communication sets out: the methods used by the manufacturer to develop SPIN software for the 1400 MWe PWRs (N4 series); the approach adopted by the IPSN to evaluate the safety software of the protection system for the N4 series of reactors.

  19. Quarterly report on Defense Nuclear Facilities Safety Board Recommendation 90-7 for the period ending December 31, 1992

    SciTech Connect

    Cash, R.J.; Dukelow, G.T.; Forbes, C.J.

    1993-03-01

    This is the seventh quarterly report on the progress of activities addressing safety issues associated with Hanford Site high-level radioactive waste tanks that contain ferrocyanide compounds. In the presence of oxidizing materials, such as nitrates or nitrites, ferrocyanide can be made to explode in the laboratory by heating it to high temperatures [above 285{degrees}C (545{degrees}F)]. In the mid 1950s approximately 140 metric tons of ferrocyanide were added to 24 underground high-level radioactive waste tanks. An implementation plan (Cash 1991) responding to the Defense Nuclear Facilities Safety Board Recommendation 90-7 (FR 1990) was issued in March 1991 describing the activities that were planned and underway to address each of the six parts of Recommendation 90-7. A revision to the original plan was transmitted to US Department of Energy by Westinghouse Hanford Company in December 1992. Milestones completed this quarter are described in this report. Contents of this report include: Introduction; Defense Nuclear Facilities Safety Board Implementation Plan Task Activities (Defense Nuclear Facilities Safety Board Recommendation for enhanced temperature measurement, Recommendation for continuous temperature monitoring, Recommendation for cover gas monitoring, Recommendation for ferrocyanide waste characterization, Recommendation for chemical reaction studies, and Recommendation for emergency response planning); Schedules; and References. All actions recommended by the Defense Nuclear Facilities Safety Board for emergency planning by Hanford Site emergency preparedness organizations have been completed.

  20. Annual report to Congress: Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, Calendar Year 1999

    SciTech Connect

    2000-02-01

    This is the tenth Annual Report to the Congress describing Department of Energy activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board (Board). The Board, an independent executive-branch agency established in 1988, provides advice and recommendations to the Secretary of Energy regarding public health and safety issues at the Department's defense nuclear facilities. The Board also reviews and evaluates the content and implementation of health and safety standards, as well as other requirements, relating to the design, construction, operation, and decommissioning of the Department's defense nuclear facilities. During 1999, Departmental activities resulted in the closure of nine Board recommendations. In addition, the Department has completed all implementation plan milestones associated with three Board recommendations. One new Board recommendation was received and accepted by the Department in 1999, and a new implementation plan is being developed to address this recommendation. The Department has also made significant progress with a number of broad-based initiatives to improve safety. These include expanded implementation of integrated safety management at field sites, opening of a repository for long-term storage of transuranic wastes, and continued progress on stabilizing excess nuclear materials to achieve significant risk reduction.

  1. The Bacillus subtilis and Bacillus halodurans Aspartyl-tRNA Synthetases Retain Recognition of tRNA(Asn).

    PubMed

    Nair, Nilendra; Raff, Hannah; Islam, Mohammed Tarek; Feen, Melanie; Garofalo, Denise M; Sheppard, Kelly

    2016-02-13

    Synthesis of asparaginyl-tRNA (Asn-tRNA(Asn)) in bacteria can be formed either by directly ligating Asn to tRNA(Asn) using an asparaginyl-tRNA synthetase (AsnRS) or by synthesizing Asn on the tRNA. In the latter two-step indirect pathway, a non-discriminating aspartyl-tRNA synthetase (ND-AspRS) attaches Asp to tRNA(Asn) and the amidotransferase GatCAB transamidates the Asp to Asn on the tRNA. GatCAB can be similarly used for Gln-tRNA(Gln) formation. Most bacteria are predicted to use only one route for Asn-tRNA(Asn) formation. Given that Bacillus halodurans and Bacillus subtilis encode AsnRS for Asn-tRNA(Asn) formation and Asn synthetases to synthesize Asn and GatCAB for Gln-tRNA(Gln) synthesis, their AspRS enzymes were thought to be specific for tRNA(Asp). However, we demonstrate that the AspRSs are non-discriminating and can be used with GatCAB to synthesize Asn. The results explain why B. subtilis with its Asn synthetase genes knocked out is still an Asn prototroph. Our phylogenetic analysis suggests that this may be common among Firmicutes and 30% of all bacteria. In addition, the phylogeny revealed that discrimination toward tRNA(Asp) by AspRS has evolved independently multiple times. The retention of the indirect pathway in B. subtilis and B. halodurans likely reflects the ancient link between Asn biosynthesis and its use in translation that enabled Asn to be added to the genetic code.

  2. Asn194Lys mutation in RVG29 peptide increases GFP transgene delivery by endocytosis to neuroblastoma and astrocyte cells.

    PubMed

    Villa-Cedillo, Sheila Adela; Rodríguez-Rocha, Humberto; Zavala-Flores, Laura Mireya; Montes-de-Oca-Luna, Roberto; García-García, Aracely; Loera-Arias, Maria de Jesus; Saucedo-Cárdenas, Odila

    2017-10-01

    A cell-penetrating peptide-based delivery system could target specific types of cells for therapeutic genes delivery. To increase the gene delivery efficiency into neuronal phenotype cells, we introduced an Asn194Lys mutation to RVG29 peptide derived from rabies virus glycoprotein and added a nuclear localization signal to enhance its nuclear import. Mutant RVG or wild-type RVG peptide, a karyophilic peptide (KP) and a plasmid encoding green fluorescent protein (pGL) were bound by electrostatic charges to form four different kinds of RVG complexes. Immunofluorescence was used to assess the gene transfection efficiency into astrocytes, oligodendrocyte precursor cells (OPCs), SH-SY5Y, HeLa and NIH/3T3 cells. The cellular uptake mechanism of RVG29 complexes was examined using endocytosis inhibitors. The mRVG29 peptide has the ability to enhance the nuclear import of plasmids. The Asn194Lys mutation in RVG29 peptide of the pGL-mRVG29 complex and the addition of KP to the pGL-RVG29-KP complex increased the capacity to deliver DNA by endocytosis in astrocytes and SH-SY5Y cells. The complexes pGL-mRVG29 and pGL-RVG29-KP have specificity for transfecting astrocytes and SH-SY5Y cells. The karyophilic capacity of this new mRVG peptide render it promising candidate to act as gene delivery vector into the brain cells. © 2017 Royal Pharmaceutical Society.

  3. Legitimating a nuclear critic: John Gofman, radiation safety, and cancer risks.

    PubMed

    Semendeferi, Ioanna

    2008-01-01

    Whether low-level ionizing radiation has an effect on humans has been a polarizing issue for the last fifty years. The epicenter of this controversy has been the validity of the linear non-threshold dose-response model, according to which any amount of radiation, however small, causes damage to human genes and health. In the late 1960s and early 1970s, the nuclear scientist and medical researcher John Gofman (1918-2007) played a pivotal role in the debate. Historical accounts have treated Gofman as a radical antinuclear scientist whose unscientific arguments put enormous political pressure on the nuclear power industry and regulatory agencies. Gofman's bitter struggle with the Atomic Energy Commission, which funded his research at Lawrence Livermore National Laboratory, partly accounts for this view. However, my analysis of Gofman's involvement in the low-level radiation debate shows how he also helped shift the focus in radiation safety from the risks of genetic damage or leukemia to somatic or cancer risks. His arguments led to the introduction of the linear non-threshold radiation model as a means of numerically estimating cancer risks. This was a watershed event in radiation-safety science and politics. Gofman's case sheds light on the process by which a scientist could secure legitimation even when his technical arguments threatened the government's interests. I conclude that it also points to an open issue in the history of antinuclear scientists, or of other politically active scientists or technology critics: treating them as critics should not preclude historians from treating them as scientists.

  4. A work observation study of nuclear medicine technologists: interruptions, resilience and implications for patient safety

    PubMed Central

    Larcos, George; Prgomet, Mirela; Georgiou, Andrew; Westbrook, Johanna

    2017-01-01

    Background Errors by nuclear medicine technologists during the preparation of radiopharmaceuticals or at other times can cause patient harm and may reflect the impact of interruptions, busy work environments and deficient systems or processes. We aimed to: (a) characterise the rate and nature of interruptions technologists experience and (b) identify strategies that support safety. Methods We performed 100 hours of observation of 11 technologists at a major public hospital and measured the proportions of time spent in eight categories of work tasks, location of task, interruption rate and type and multitasking (tasks conducted in parallel). We catalogued specific safety-oriented strategies used by technologists. Results Technologists completed 5227 tasks and experienced 569 interruptions (mean, 4.5 times per hour; 95% CI 4.1 to 4.9). The highest interruption rate occurred when technologists were in transit between rooms (10.3 per hour (95% CI 8.3 to 12.5)). Interruptions during radiopharmaceutical preparation occurred a mean of 4.4 times per hour (95% CI 3.3 to 5.6). Most (n=426) tasks were interrupted once only and all tasks were resumed after interruption. Multitasking occurred 16.6% of the time. At least some interruptions were initiated by other technologists to convey important information and/or to render assistance. Technologists employed a variety of verbal and non-verbal strategies in all work areas (notably in the hot-lab) to minimise the impact of interruptions and optimise the safe conduct of procedures. Although most were due to individual choices, some strategies reflected overt or subliminal departmental policy. Conclusions Some interruptions appear beneficial. Technologists' self-initiated strategies to support safe work practices appear to be an important element in supporting a resilient work environment in nuclear medicine. PMID:27707869

  5. General-purpose heat source project and space nuclear safety and fuels program. Progress report

    SciTech Connect

    Maraman, W.J.

    1980-02-01

    Studies related to the use of /sup 238/PuO/sub 2/ in radioisotopic power systems carried out for the Advanced Nuclear Systems and Projects Division of LASL are presented. The three programs involved are: general-purpose heat source development; space nuclear safety; and fuels program. Three impact tests were conducted to evaluate the effects of a high temperature reentry pulse and the use of CBCF on impact performance. Additionally, two /sup 238/PuO/sub 2/ pellets were encapsulated in Ir-0.3% W for impact testing. Results of the clad development test and vent testing are noted. Results of the environmental tests are summarized. Progress on the Stirling isotope power systems test and the status of the improved MHW tests are indicated. The examination of the impact failure of the iridium shell of MHFT-65 at a fuel pass-through continued. A test plan was written for vibration testing of the assembled light-weight radioisotopic heater unit. Progress on fuel processing is reported.

  6. Assessment of modular construction for safety-related structures at advanced nuclear power plants

    SciTech Connect

    Braverman, J.; Morante, R.; Hofmayer, C.

    1997-03-01

    Modular construction techniques have been successfully used in a number of industries, both domestically and internationally. Recently, the use of structural modules has been proposed for advanced nuclear power plants. The objective in utilizing modular construction is to reduce the construction schedule, reduce construction costs, and improve the quality of construction. This report documents the results of a program which evaluated the proposed use of modular construction for safety-related structures in advanced nuclear power plant designs. The program included review of current modular construction technology, development of licensing review criteria for modular construction, and initial validation of currently available analytical techniques applied to concrete-filled steel structural modules. The program was conducted in three phases. The objective of the first phase was to identify the technical issues and the need for further study in order to support NRC licensing review activities. The two key findings were the need for supplementary review criteria to augment the Standard Review Plan and the need for verified design/analysis methodology for unique types of modules, such as the concrete-filled steel module. In the second phase of this program, Modular Construction Review Criteria were developed to provide guidance for licensing reviews. In the third phase, an analysis effort was conducted to determine if currently available finite element analysis techniques can be used to predict the response of concrete-filled steel modules.

  7. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.; Baker, L. Jr.

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and can be discharged from the reactor core. The invention provides a porous bed of sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  8. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker Jr., Louis

    1986-07-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  9. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOEpatents

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker, Jr., Louis

    1986-01-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  10. Stochastic sampling method with MCNPX for nuclear data uncertainty propagation in criticality safety applications

    SciTech Connect

    Zhu, T.; Vasiliev, A.; Wieselquist, W.; Ferroukhi, H.

    2012-07-01

    In the domain of criticality safety, the efficient propagation of uncertainty in nuclear data to uncertainty in k{sub eff} is an important area of current research. In this paper, a method based on stochastic sampling is presented for uncertainty propagation in MCNPX calculations. To that aim, the nuclear data (i.e. cross sections) are assumed to have a multivariate normal distribution and simple random sampling is performed following this presumed probability distribution. A verification of the developed stochastic sampling procedure with MCNPX is then conducted using the {sup 239}Pu Jezebel experiment as well as the PB-2 BWR and TMI-1 PWR pin cell models from the Uncertainty Analysis in Modeling (UAM) exercises. For the Jezebel case, it is found that the developed stochastic sampling approach predicts similar k{sub eff} uncertainties compared to conventional sensitivity and uncertainty methods. For the UAM models, slightly lower uncertainties are obtained when comparing to existing preliminary results. Further details of these verification studies are discussed and directions for future work are outlined. (authors)

  11. Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

    NASA Astrophysics Data System (ADS)

    Yi, Ping; Wang, Qingkang; Kong, Xianjing

    2017-01-01

    The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel (PCCV) model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

  12. Aging of turbine drives for safety-related pumps in nuclear power plants

    SciTech Connect

    Cox, D.F.

    1995-06-01

    This study was performed to examine the relationship between time-dependent degradation and current industry practices in the areas of maintenance, surveillance, and operation of steam turbine drives for safety-related pumps. These pumps are located in the Auxiliary Feedwater (AFW) system for pressurized-water reactor plants and in the Reactor Core Isolation Cooling and High-Pressure Coolant Injection systems for boiling-water reactor plants. This research has been conducted by examination of failure data in the Nuclear Plant Reliability Data System, review of Licensee Event Reports, discussion of problems with operating plant personnel, and personal observation. The reported failure data were reviewed to determine the cause of the event and the method of discovery. Based on the research results, attempts have been made to determine the predictability of failures and possible preventive measures that may be implemented. Findings in a recent study of AFW systems indicate that the turbine drive is the single largest contributor to AFW system degradation. However, examination of the data shows that the turbine itself is a reliable piece of equipment with a good service record. Most of the problems documented are the result of problems with the turbine controls and the mechanical overspeed trip mechanism; these apparently stem from three major causes which are discussed in the text. Recent improvements in maintenance practices and procedures, combined with a stabilization of the design, have led to improved performance resulting in a reliable safety-related component. However, these improvements have not been universally implemented.

  13. Seismic performance assessment of base-isolated safety-related nuclear structures

    USGS Publications Warehouse

    Huang, Y.-N.; Whittaker, A.S.; Luco, N.

    2010-01-01

    Seismic or base isolation is a proven technology for reducing the effects of earthquake shaking on buildings, bridges and infrastructure. The benefit of base isolation has been presented in terms of reduced accelerations and drifts on superstructure components but never quantified in terms of either a percentage reduction in seismic loss (or percentage increase in safety) or the probability of an unacceptable performance. Herein, we quantify the benefits of base isolation in terms of increased safety (or smaller loss) by comparing the safety of a sample conventional and base-isolated nuclear power plant (NPP) located in the Eastern U.S. Scenario- and time-based assessments are performed using a new methodology. Three base isolation systems are considered, namely, (1) Friction Pendulum??? bearings, (2) lead-rubber bearings and (3) low-damping rubber bearings together with linear viscous dampers. Unacceptable performance is defined by the failure of key secondary systems because these systems represent much of the investment in a new build power plant and ensure the safe operation of the plant. For the scenario-based assessments, the probability of unacceptable performance is computed for an earthquake with a magnitude of 5.3 at a distance 7.5 km from the plant. For the time-based assessments, the annual frequency of unacceptable performance is computed considering all potential earthquakes that may occur. For both assessments, the implementation of base isolation reduces the probability of unacceptable performance by approximately four orders of magnitude for the same NPP superstructure and secondary systems. The increase in NPP construction cost associated with the installation of seismic isolators can be offset by substantially reducing the required seismic strength of secondary components and systems and potentially eliminating the need to seismically qualify many secondary components and systems. ?? 2010 John Wiley & Sons, Ltd.

  14. Additional Studies of the Criticality Safety of Failed Used Nuclear Fuel

    SciTech Connect

    Marshall, William BJ J; Wagner, John C

    2013-01-01

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories and specific configurations were evaluated to understand trends and quantify the consequences of worst-case potential reconfiguration progressions. These results will be summarized here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g., >20% keff). It can be concluded that the consequences of credible fuel failure configurations from ES or transportation following ES are manageable (e.g., <5% keff). The current work expands on these efforts and examines some modified scenarios and modified approaches to investigate the effectiveness of some techniques for reducing the calculated increase in keff. The areas included here are more realistic modeling of some assembly types and the effect of reconfiguration of some assemblies in the storage and transportation canister.

  15. Fault displacement hazard assessment for nuclear installations based on IAEA safety standards

    NASA Astrophysics Data System (ADS)

    Fukushima, Y.

    2016-12-01

    In the IAEA Safety NS-R-3, surface fault displacement hazard assessment (FDHA) is required for the siting of nuclear installations. If any capable faults exist in the candidate site, IAEA recommends the consideration of alternative sites. However, due to the progress in palaeoseismological investigations, capable faults may be found in existing site. In such a case, IAEA recommends to evaluate the safety using probabilistic FDHA (PFDHA), which is an empirical approach based on still quite limited database. Therefore a basic and crucial improvement is to increase the database. In 2015, IAEA produced a TecDoc-1767 on Palaeoseismology as a reference for the identification of capable faults. Another IAEA Safety Report 85 on ground motion simulation based on fault rupture modelling provides an annex introducing recent PFDHAs and fault displacement simulation methodologies. The IAEA expanded the project of FDHA for the probabilistic approach and the physics based fault rupture modelling. The first approach needs a refinement of the empirical methods by building a world wide database, and the second approach needs to shift from kinematic to the dynamic scheme. Both approaches can complement each other, since simulated displacement can fill the gap of a sparse database and geological observations can be useful to calibrate the simulations. The IAEA already supported a workshop in October 2015 to discuss the existing databases with the aim of creating a common worldwide database. A consensus of a unified database was reached. The next milestone is to fill the database with as many fault rupture data sets as possible. Another IAEA work group had a WS in November 2015 to discuss the state-of-the-art PFDHA as well as simulation methodologies. Two groups jointed a consultancy meeting in February 2016, shared information, identified issues, discussed goals and outputs, and scheduled future meetings. Now we may aim at coordinating activities for the whole FDHA tasks jointly.

  16. Landscape modeling for dose calculations in the safety assessment of a repository for spent nuclear fuel

    SciTech Connect

    Lindborg, Tobias; Kautsky, Ulrik; Brydsten, Lars

    2007-07-01

    The Swedish Nuclear Fuel and Waste Management Co.,(SKB), pursues site investigations for the final repository for spent nuclear fuel at two sites in the south eastern part of Sweden, the Forsmark- and the Laxemar site. Data from the two site investigations are used to build site descriptive models of the areas. These models describe the bedrock and surface system properties important for designing the repository, the environmental impact assessment, and the long-term safety, i.e. up to 100,000 years, in a safety assessment. In this paper we discuss the methodology, and the interim results for, the landscape model, used in the safety assessment to populate the Forsmark site in the numerical dose models. The landscape model is built upon ecosystem types, e.g. a lake or a mire, (Biosphere Objects) that are connected in the landscape via surface hydrology. Each of the objects have a unique set of properties derived from the site description. The objects are identified by flow transport modeling, giving discharge points at the surface for all possible flow paths from the hypothetical repository in the bedrock. The landscape development is followed through time by using long-term processes e.g. shoreline displacement and sedimentation. The final landscape model consists of a number of maps for each chosen time period and a table of properties that describe the individual objects which constitutes the landscape. The results show a landscape that change over time during 20,000 years. The time period used in the model equals the present interglacial and can be used as an analogue for a future interglacial. Historically, the model area was covered by sea, and then gradually changes into a coastal area and, in the future, into a terrestrial inland landscape. Different ecosystem types are present during the landscape development, e.g. sea, lakes, agricultural areas, forest and wetlands (mire). The biosphere objects may switch from one ecosystem type to another during the

  17. Peer review of the Barselina Level 1 probabilistic safety assessment of the Ignalina Nuclear Power Plant, Unit 2

    SciTech Connect

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

    1995-01-01

    The Barselina Project is a Swedish-funded, cooperative effort among Lithuania, Russia and Sweden to transfer Western probabilistic safety assessment (PSA) methodology to the designers/operators of Ignalina Nuclear Power Plant (INPP). The overall goal is to use the PSA as a tool for assessing plant operational safety. The INPP is a two-unit, Former Soviet Union-designed nuclear facility located in Lithuania. The results of this PSA will ultimately be used to identify plant-specific improvements in system design and the conduct of facility operations, allowing improved operational safety. Pacific Northwest Laboratory (PNL) was asked to perform an independent expert peer review of the Barselina PSA. This report documents the findings of this review. This review, financed with nuclear safety assistance funds through the US Agency for International Development (USAID) and the US Department of Energy (DOE), satisfies Task II of the PNL peer review of the Barselina project. The objective is to provide an independent, in-proce ss examination of the Barselina Level 1 PSA of Ignalina Nuclear Power Plant, Unit 2. The review consisted of an investigation of the project documentation, interviews, and extensive discussions with the PSA staff during critical stages of the project. PNL assessed the readability, completeness, consistency, validity, and applicability of the PSA. The major aspects explored were its purpose, major assumptions, analysis/modeling, results, and interpretation. It was not within the scope of this review to perform plant walkdowns or to review material other than the PSA documentation.

  18. Safety of Soviet-designed nuclear reactors: a general assessment and international efforts to upgrade their safety

    NASA Astrophysics Data System (ADS)

    Speis, Themis P.

    1995-03-01

    This paper provides a general description of the various Soviet-designed reactor types operating in the countries of the former Soviet Union (FSU) and Eastern and Central Europe and the various safety issues associated with each reactor type, in terms of their design and operation. It also provides a general safety assessment and the technical potential for safety improvement for each reactor type (including potential short-term and longer-term actions). Other issues discussed, including their effect on safety, include organizational problems, lack of well established regulatory organizations, and the changing socio-political and economic situation. The condition of the Chernobyl sarcophagus and safety concerns related to it are addressed. Finally, international efforts and programs underway, including those originating in the USA, to provide financial and technical assistance to secure safety improvements and encourage indigenous improvements are noted.

  19. Concentration of Actinides in Plant Mounds at Safety Test Nuclear Sites in Nevada

    SciTech Connect

    David S. Shafer; Jenna Gommes

    2008-09-15

    Plant mounds or blow-sand mounds are accumulations of soil particles and plant debris around large shrubs and are common features in deserts in the southwestern United States. Believed to be an important factor in their formation, the shrubs create surface roughness that causes wind-suspended particles to be deposited and resist further suspension. Shrub mounds occur in some plant communities on the Nevada Test Site, the Nevada Test and Training Range (NTTR), and Tonopah Test Range (TTR), including areas of surface soil contamination from past nuclear testing. In the 1970s as part of early studies to understand properties of actinides in the environment, the Nevada Applied Ecology Group (NAEG) examined the accumulation of isotopes of Pu, {sup 241}Am, and U in plant mounds at safety test sites. The NAEG studies found concentrations of these contaminants to be greater in shrub mounds than in the surrounding areas of desert pavement. For example, at Project 57 on the NTTR, it was estimated that 15 percent of the radionuclide inventory of the site was associated with shrub mounds, which accounted for 17 percent of the surface area of the site, a ratio of inventory to area of 0.85. At Clean Slate III at the TTR, 29 percent of the inventory was associated with approximately 32 percent of the site covered by shrub mounds, a ratio of 0.91. While the total inventory of radionuclides in intershrub areas was greater, the ratio of radionuclide inventory to area was 0.40 and 0.38, respectively, at the two sites. The comparison between the shrub mounds and adjacent desert pavement areas was made for only the top 5 cm since radionuclides at safety test sites are concentrated in the top 5 cm of intershrub areas. Not accounting for radionuclides associated with the shrub mounds would cause the inventory of contaminants and potential exposure to be underestimated. As part of its Environmental Restoration Soils Subproject, the U.S. Department of Energy (DOE), National Nuclear

  20. The Development, Content, Design, and Conduct of the 2011 Piloted US DOE Nuclear Criticality Safety Program Criticality Safety Engineering Training and Education Project

    SciTech Connect

    Hopper, Calvin Mitchell

    2011-01-01

    In May 1973 the University of New Mexico conducted the first nationwide criticality safety training and education week-long short course for nuclear criticality safety engineers. Subsequent to that course, the Los Alamos Critical Experiments Facility (LACEF) developed very successful 'hands-on' subcritical and critical training programs for operators, supervisors, and engineering staff. Since the inception of the US Department of Energy (DOE) Nuclear Criticality Technology and Safety Project (NCT&SP) in 1983, the DOE has stimulated contractor facilities and laboratories to collaborate in the furthering of nuclear criticality as a discipline. That effort included the education and training of nuclear criticality safety engineers (NCSEs). In 1985 a textbook was written that established a path toward formalizing education and training for NCSEs. Though the NCT&SP went through a brief hiatus from 1990 to 1992, other DOE-supported programs were evolving to the benefit of NCSE training and education. In 1993 the DOE established a Nuclear Criticality Safety Program (NCSP) and undertook a comprehensive development effort to expand the extant LACEF 'hands-on' course specifically for the education and training of NCSEs. That successful education and training was interrupted in 2006 for the closing of the LACEF and the accompanying movement of materials and critical experiment machines to the Nevada Test Site. Prior to that closing, the Lawrence Livermore National Laboratory (LLNL) was commissioned by the US DOE NCSP to establish an independent hands-on NCSE subcritical education and training course. The course provided an interim transition for the establishment of a reinvigorated and expanded two-week NCSE education and training program in 2011. The 2011 piloted two-week course was coordinated by the Oak Ridge National Laboratory (ORNL) and jointly conducted by the Los Alamos National Laboratory (LANL) classroom education and facility training, the Sandia National

  1. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    SciTech Connect

    Korsah, K.; Wood, R.T.; Hassan, M.; Tanaka, T.J.

    1998-01-01

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.

  2. The effect of leader communication style on safety-conscious work environments at domestic nuclear power plants

    NASA Astrophysics Data System (ADS)

    Goldberg, Edward Michael

    Public risk from unsafe nuclear power plant operations increases when plant workers are reluctant to raise issues and concerns. The effect of leader communication style on the safety-conscious work environment (SCWE) at domestic nuclear power plants was evaluated using a descriptive quantitative research study. A sample of 379 plant employees was surveyed to determine leader communication style elements that foster SCWE. The results reveal that leader communication style significantly affects a safety-conscious work environment. Specific attributes such as wit, articulation, self-disclosure, and social composure, confirmation, and experience, were proven to directly affect worker's likelihood to raise issues and concerns. The direct effect of leader, communication style on safe plant operations and the communication actions leaders can take to improve the safety of those operations is discussed.

  3. Operational safety enhancement of Soviet-designed nuclear reactors via development of nuclear power plant simulators and transfer of related technology

    SciTech Connect

    Kohut, P.; Epel, L.G.; Tutu, N.K.

    1998-08-01

    The US Department of Energy (DOE), under the US government`s International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper.

  4. Contributions to nuclear safety and radiation technologies in Ukraine by the Science and Technology Center in Ukraine (STCU)

    NASA Astrophysics Data System (ADS)

    Taranenko, L.; Janouch, F.; Owsiacki, L.

    2001-06-01

    This paper presents Science and Technology Center in Ukraine (STCU) activities devoted to furthering nuclear and radiation safety, which is a prioritized STCU area. The STCU, an intergovernmental organization with the principle objective of non-proliferation, administers financial support from the USA, Canada, and the EU to Ukrainian projects in various scientific and technological areas; coordinates projects; and promotes the integration of Ukrainian scientists into the international scientific community, including involving western collaborators. The paper focuses on STCU's largest project to date "Program Supporting Y2K Readiness at Ukrainian NPPs" initiated in April 1999 and designed to address possible Y2K readiness problems at 14 Ukrainian nuclear reactors. Other presented projects demonstrate a wide diversity of supported directions in the fields of nuclear and radiation safety, including reactor material improvement ("Improved Zirconium-Based Elements for Nuclear Reactors"), information technologies for nuclear industries ("Ukrainian Nuclear Data Bank in Slavutich"), and radiation health science ("Diagnostics and Treatment of Radiation-Induced Injuries of Human Biopolymers").

  5. Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code

    SciTech Connect

    Turner, John A; Clarno, Kevin T; Hansen, Glen A

    2009-09-01

    Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied more on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel

  6. New neutron cross-section measurements on {sup 19}F, {sup 39,41}K, {sup 55}Mn, and {sup 103}Rh for improved nuclear criticality safety

    SciTech Connect

    Guber, K. H.; Leal, L. C.; Sayer, R. O.; Koehler, P. E.; Wiarda, D.; Valentine, T. E.; Derrien, H.; Harvey, J. A.; Kopecky, S.; Siegler, P.; Schillebeeckx, P.; Wynants, R.; Ivanov, I.; Borella, A.

    2006-07-01

    A series of new measurements has been undertaken in response to deficiencies identified in nuclear data libraries of crucial importance to the Nuclear Criticality Safety Program as well as for burnup credit studies involving the transportation of spent nuclear fuel. New data and evaluations including covariances are required for several stable fission products as well as for materials found in mixtures with uranium. (authors)

  7. Conceptual Software Reliability Prediction Models for Nuclear Power Plant Safety Systems

    SciTech Connect

    Johnson, G.; Lawrence, D.; Yu, H.

    2000-04-03

    The objective of this project is to develop a method to predict the potential reliability of software to be used in a digital system instrumentation and control system. The reliability prediction is to make use of existing measures of software reliability such as those described in IEEE Std 982 and 982.2. This prediction must be of sufficient accuracy to provide a value for uncertainty that could be used in a nuclear power plant probabilistic risk assessment (PRA). For the purposes of the project, reliability was defined to be the probability that the digital system will successfully perform its intended safety function (for the distribution of conditions under which it is expected to respond) upon demand with no unintended functions that might affect system safety. The ultimate objective is to use the identified measures to develop a method for predicting the potential quantitative reliability of a digital system. The reliability prediction models proposed in this report are conceptual in nature. That is, possible prediction techniques are proposed and trial models are built, but in order to become a useful tool for predicting reliability, the models must be tested, modified according to the results, and validated. Using methods outlined by this project, models could be constructed to develop reliability estimates for elements of software systems. This would require careful review and refinement of the models, development of model parameters from actual experience data or expert elicitation, and careful validation. By combining these reliability estimates (generated from the validated models for the constituent parts) in structural software models, the reliability of the software system could then be predicted. Modeling digital system reliability will also require that methods be developed for combining reliability estimates for hardware and software. System structural models must also be developed in order to predict system reliability based upon the reliability

  8. [A questionnaire about radiation safety management of the draining-water system at nuclear medicine facilities].

    PubMed

    Shizukuishi, Kazuya; Watanabe, Hiroshi; Narita, Hiroto; Kanaya, Shinichi; Kobayashi, Kazumi; Yamamoto, Tetsuo; Tsukada, Masaru; Iwanaga, Tetsuo; Ikebuchi, Shuji; Kusama, Keiji; Tanaka, Mamoru; Namiki, Norio; Fuiimura, Youko; Horikoshi, Akiko; Inoue, Tomio; Kusakabe, Kiyoko

    2004-05-01

    We conducted a questionnaire survey about radiation-safety management condition in Japanese nuclear medicine facilities to make materials of proposition for more reasonable management of medical radioactive waste. We distributed a questionnaire to institutions equipped with Nuclear Medicine facilities. Of 1,125 institutions, 642 institutes (52.8%) returned effective answers. The questionnaire covered the following areas: 1) scale of an institution, 2) presence of enforcement of radiotherapy, 3) system of a tank, 4) size and number of each tank, 5) a form of draining-water system, 6) a displacement in a radioactive rays management area, 7) a measurement method of the concentration of medical radioactive waste in draining water system, 8) planned and used quantity of radioisotopes for medical examination and treatment, 9) an average displacement of hospital for one month. In most institutions, a ratio of dose limitation of radioisotope in draining-water system was less than 1.0, defined as an upper limitation in ordinance. In 499 hospitals without facilities of hospitalization for unsealed radioisotope therapy, 473 hospitals reported that sum of ratios of dose limits in a draining-water system was less than 1.0. It was calculated by used dose of radioisotope and monthly displacement from hospital, on the premise that all used radioisotope entered in the general draining-water system. When a drainage including radioactivity from a controlled area join with that from other area before it flows out of a institution, it may be diluted and its radioactive concentration should be less than its upper limitation defined in the rule. Especially, in all institutions with a monthly displacement of more than 25,000 m3, the sum of ratio of the concentration of each radionuclide to the concentration limit dose calculated by used dose of radioisotope, indicated less than 1.0.

  9. The ORSphere Benchmark Evaluation and Its Potential Impact on Nuclear Criticality Safety

    SciTech Connect

    John D. Bess; Margaret A. Marshall; J. Blair Briggs

    2013-10-01

    In the early 1970’s, critical experiments using an unreflected metal sphere of highly enriched uranium (HEU) were performed with the focus to provide a “very accurate description…as an ideal benchmark for calculational methods and cross-section data files.” Two near-critical configurations of the Oak Ridge Sphere (ORSphere) were evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results from those benchmark experiments were then compared with additional unmoderated and unreflected HEU metal benchmark experiment configurations currently found in the ICSBEP Handbook. For basic geometries (spheres, cylinders, and slabs) the eigenvalues calculated using MCNP5 and ENDF/B-VII.0 were within 3 of their respective benchmark values. There appears to be generally a good agreement between calculated and benchmark values for spherical and slab geometry systems. Cylindrical geometry configurations tended to calculate low, including more complex bare HEU metal systems containing cylinders. The ORSphere experiments do not calculate within their 1s uncertainty and there is a possibility that the effect of the measured uncertainties for the GODIVA I benchmark may need reevaluated. There is significant scatter in the calculations for the highly-correlated ORCEF cylinder experiments, which are constructed from close-fitting HEU discs and annuli. Selection of a nuclear data library can have a larger impact on calculated eigenvalue results than the variation found within calculations of a given experimental series, such as the ORCEF cylinders, using a single nuclear data set.

  10. Mathematical aspects of assessing extreme events for the safety of nuclear plants

    NASA Astrophysics Data System (ADS)

    Potempski, Slawomir; Borysiewicz, Mieczyslaw

    2015-04-01

    In the paper the review of mathematical methodologies applied for assessing low frequencies of rare natural events like earthquakes, tsunamis, hurricanes or tornadoes, floods (in particular flash floods and surge storms), lightning, solar flares, etc., will be given in the perspective of the safety assessment of nuclear plants. The statistical methods are usually based on the extreme value theory, which deals with the analysis of extreme deviation from the median (or the mean). In this respect application of various mathematical tools can be useful, like: the extreme value theorem of Fisher-Tippett-Gnedenko leading to possible choices of general extreme value distributions, or the Pickands-Balkema-de Haan theorem for tail fitting, or the methods related to large deviation theory. In the paper the most important stochastic distributions relevant for performing rare events statistical analysis will be presented. This concerns, for example, the analysis of the data with the annual extreme values (maxima - "Annual Maxima Series" or minima), or the peak values, exceeding given thresholds at some periods of interest ("Peak Over Threshold"), or the estimation of the size of exceedance. Despite of the fact that there is a lack of sufficient statistical data directly containing rare events, in some cases it is still possible to extract useful information from existing larger data sets. As an example one can consider some data sets available from the web sites for floods, earthquakes or generally natural hazards. Some aspects of such data sets will be also presented taking into account their usefulness for the practical assessment of risk for nuclear power plants coming from extreme weather conditions.

  11. Neutronics and safety analysis of pellet bed reactor for nuclear thermal propulsion

    SciTech Connect

    Morley, N.J.; El-Genk, M.S. )

    1993-01-15

    The Pellet Bed Reactor for Nuclear Thermal Propulsion is modeled using the TWODANT discrete ordinance code to determine a reactor point design based on the selection of a fuel fraction in and a diameter of the pellets, dimensions of the reactor core, maximum fuel temperature, and sub-criticality during a water flooding accident. A total excess reactivity of approximately $1.25 (or k[sub eff] of 1.01), an order of magnitude higher than that estimated at BOM for 15 hours of full power, steady-state operation of the PeBR, is considered. Besides calculating the dimensions of the reactor core to satisfy the excess reactivity at BOM, the results of the neutronics calculations include estimates of the radial and axial fission power density profiles in the PeBR core. These results, in conjunction with a 1-D, steady-state thermal hydraulics analysis are used to select the operation and design characteristics of the PeBR point design, namely: (a) core radius and height of 38.4 cm and 120 cm, respectively, (b) pellet matrix fraction of 0.5, (c) total reactor mass of 3500 kg, excluding those of the radiation shield, the propulsion nozzle, external structure for the propellant flow into the core, and the drive mechanisms of the control drums in the radial reflector, (d) power density of 10 and 15 MW/l for a reactor thermal power of 1000 MW and 1500 MW, submersion calculations show that with all safety rods removed from the core, the 16 control drums are insufficient to maintain the reactor sub-critical. However, when the 8, B[sub 4]C safety rods are inserted into the reactor, it is possible to maintain the submerged PeBR point design $7.5 sub-critical (k[sub eff]=0.94).

  12. Neutronics and safety analysis of pellet bed reactor for nuclear thermal propulsion

    NASA Astrophysics Data System (ADS)

    Morley, Nicholas J.; El-Genk, Mohamed S.

    1993-01-01

    The Pellet Bed Reactor for Nuclear Thermal Propulsion is modeled using the TWODANT discrete ordinance code to determine a reactor point design based on the selection of a fuel fraction in and a diameter of the pellets, dimensions of the reactor core, maximum fuel temperature, and sub-criticality during a water flooding accident. A total excess reactivity of approximately 1.25 (or keff of 1.01), an order of magnitude higher than that estimated at BOM for 15 hours of full power, steady-state operation of the PeBR, is considered. Besides calculating the dimensions of the reactor core to satisfy the excess reactivity at BOM, the results of the neutronics calculations include estimates of the radial and axial fission power density profiles in the PeBR core. These results, in conjunction with a 1-D, steady-state thermal hydraulics analysis are used to select the operation and design characteristics of the PeBR point design, namely: (a) core radius and height of 38.4 cm and 120 cm, respectively, (b) pellet matrix fraction of 0.5, (c) total reactor mass of 3500 kg, excluding those of the radiation shield, the propulsion nozzle, external structure for the propellant flow into the core, and the drive mechanisms of the control drums in the radial reflector, (d) power density of 10 and 15 MW/l for a reactor thermal power of 1000 MW and 1500 MW, submersion calculations show that with all safety rods removed from the core, the 16 control drums are insufficient to maintain the reactor sub-critical. However, when the 8, B4C safety rods are inserted into the reactor, it is possible to maintain the submerged PeBR point design 7.5 sub-critical (keff=0.94).

  13. Generic requirements specification for qualifying a commercially available PLC for safety-related applications in nuclear power plants. Final report

    SciTech Connect

    Ostenso, A.; May, R.

    1996-12-01

    This is a specification for qualifying a commercially available PLC for application to safety systems in nuclear power plants. The specifications are suitable for evaluating a particular PLC product line as a platform for safety-related applications, establishing a suitable qualification test program, and confirming that the manufacturer has a quality assurance program that is adequate for safety-related applications or is sufficiently complete that, with a reasonable set of compensatory actions, it can be brought into conformance. The specification includes requirements for: (1) quality assurance measures applied to the qualification activities, (2) documentation to support the qualification, and (3) documentation to provide the information needed for applying the qualified PLC platform to a specific application. The specifications are designed to encompass a broad range of safety applications; however, qualifying a particular platform for a different range of applications can be accomplished by appropriate adjustments to the requirements.

  14. Allocating resources and building confidence in public-safety decisions for nuclear waste sites

    SciTech Connect

    Lew, K L; Wilder, D G

    1999-05-21

    There are three basic ways to protect the public from the hazards of exposure to radionuclides in nuclear waste: completely contain the waste; limit the rate at which radionuclides are released; and, once radionuclides are released, minimize their impact by reducing concentrations and retarding transport. A geologic repository system that implements all three provides maximum protection for the public: if one element fails, the others serve to protect. This is ''defense-in-depth.'' Demonstrating confidence in the ability of a designed system to provide the requisite safety to the public must rely on a combination of the following aspects relating to engineered and natural system components: 1 Knowledge or understanding of properties and processes 2 Uniformity of (or ability to understand or control) the range of variability associated with each component 3 Experience over time This paper proposes a tool based on defining a ''confidence region'' determined by these three essential aspects of confidence. The defense-in-depth decision-making tool described identifies the portion of the ultimate confidence region that is not well demonstrated and indicates where there is potential for changing a specific component's confidence region, therefore providing in-formation for decisions on emphasis--either for demonstrating performance or for focusing on further studies. The US Yucca Mountain Site Characterization Project (YMP), wherein Yucca Mountain is being investigated as a potential site for a nuclear waste repository, and the Swedish geologic repository studies are used as examples of this tool. of protective or operating components such that failure of a single component does not by itself lead to system failure. The greater the exposure to loss, the greater the requirements for design margins (the margin of conservatism associated with the fabrication and operation of important components in complex engineering projects) or for compensation by defense-in-depth. Thus

  15. Recommended electromagnetic operating envelopes for safety-related I and C systems in nuclear power plants: Draft report for comment

    SciTech Connect

    Ewing, P.D.; Wood, R.T.

    1997-12-01

    This document presents recommendations for electromagnetic operating envelopes to augment test criteria and test methods addressing electromagnetic interference (EMI), radio-frequency interference (RFI), and power surges that are applicable to safety-related instrumentation and control (I and C) systems in nuclear power plants. The Oak Ridge National Laboratory (ORNL) was engaged by the US Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research to assist in developing the technical basis for regulatory guidance on EMI/RFI immunity and power surge withstand capability (SWC). Previous research has provided recommendations on electromagnetic compatibility (EMC) design and installation practices, endorsement of EMI/RFI immunity and SWC test criteria and test methods, and determination of ambient electromagnetic conditions at nuclear power plants. The present research involves development of recommended electromagnetic envelopes that are applicable to nuclear power plant locations where safety-related I and C systems either are or may be installed. These recommended envelopes establish both emissions criteria and the levels of radiated and conducted interference that I and C systems should be able to withstand without upset or malfunction. The EMI/RFI operating envelopes are derived from conditions in comparable military environments and are confirmed by comparison with the nuclear power plant electromagnetic environment based on measured plant emissions profiles. Detailed information on specific power surge conditions in nuclear power plants is not available, so industrial guidance on representative surge characteristics for susceptibility testing is adopted. An engineering assessment of the power surge environment in nuclear power plants leads to the recommendation of operating envelopes based on location categories and exposure levels defined in IEEE Std C62.41-1991, IEEE Recommended Practice on Surge Voltages in Low-Voltage AC Power Circuits.

  16. French policy for managing the post-accident phase of a nuclear accident.

    PubMed

    Gallay, F; Godet, J L; Niel, J C

    2015-06-01

    In 2005, at the request of the French Government, the Nuclear Safety Authority (ASN) established a Steering Committee for the Management of the Post-Accident Phase of a Nuclear Accident or a Radiological Emergency, with the objective of establishing a policy framework. Under the supervision of ASN, this Committee, involving several tens of experts from different backgrounds (e.g. relevant ministerial offices, expert agencies, local information commissions around nuclear installations, non-governmental organisations, elected officials, licensees, and international experts), developed a number of recommendations over a 7-year period. First published in November 2012, these recommendations cover the immediate post-emergency situation, and the transition and longer-term periods of the post-accident phase in the case of medium-scale nuclear accidents causing short-term radioactive release (less than 24 h) that might occur at French nuclear facilities. They also apply to actions to be undertaken in the event of accidents during the transportation of radioactive materials. These recommendations are an important first step in preparation for the management of a post-accident situation in France in the case of a nuclear accident. © The Chartered Institution of Building Services Engineers 2014.

  17. Understanding Earthquake Processes in the Central and Eastern US and Implications for Nuclear Reactor Safety

    NASA Astrophysics Data System (ADS)

    Seber, D.; Tabatabai, S.

    2012-12-01

    All of the early site permits and new reactor licensing applications, which have been submitted to the U.S. Nuclear Regulatory Commission (U.S. NRC), are located in the Central and Eastern United States (CEUS). Furthermore, among the 104 commercial nuclear power plants (NPPs) already licensed to operate in the US, 96 are located in the CEUS. While there are many considerations in siting commercial NPPs, the perceived lower seismic hazard in the CEUS compared to the Western United States is one of the reasons why the majority of operating and potential future nuclear reactors are located in the CEUS. However, one important criterion used in the licensing and safe operation of a nuclear power plant is its seismic design basis, which establishes the plant's ability to withstand ground motions produced by moderate- to large-sized earthquakes without suffering any damage to its critical safety related structures, systems, and components. The seismic design basis for a NPP is site specific and determined using up-to-date knowledge and information about seismic sources surrounding the site and seismic wave propagation characteristics. Therefore, an in-depth understanding of the processes generating earthquakes (tectonic or man-made) and the seismic wave propagation characteristics in the CEUS is crucial. The U.S. NRC's seismic review process for evaluating new reactor siting applications heavily relies upon up-to-date scientific knowledge of seismic sources within at least 320 km of a proposed site. However, the availability of up-to-date knowledge and information about potential seismic sources in low-seismicity regions is limited and relevant data are sparse. Recently, the NRC participated in a joint effort to develop new seismic source models to be used in the CEUS seismic hazard studies for nuclear facilities. In addition, efforts are underway to better understand the seismic potential of the Eastern Tennessee Seismic Zone. While very large and successful scientific

  18. 77 FR 20853 - Entergy Nuclear Operations, Inc.; Establishment of Atomic Safety and Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-06

    ... over the following proceeding: Entergy Nuclear Operations, Inc. (Pilgrim Nuclear Power Station) A... Association and by Pilgrim Watch seeking leave to reopen the record and request a hearing. Petitioners filed a... application from Entergy Nuclear Operations, Inc. to renew the current operating license for Pilgrim Nuclear...

  19. Structural Safety Analysis Based on Seismic Service Conditions for Butterfly Valves in a Nuclear Power Plant

    PubMed Central

    Han, Sang-Uk; Ahn, Dae-Gyun; Lee, Myeong-Gon

    2014-01-01

    The structural integrity of valves that are used to control cooling waters in the primary coolant loop that prevents boiling within the reactor in a nuclear power plant must be capable of withstanding earthquakes or other dangerous situations. In this study, numerical analyses using a finite element method, that is, static and dynamic analyses according to the rigid or flexible characteristics of the dynamic properties of a 200A butterfly valve, were performed according to the KEPIC MFA. An experimental vibration test was also carried out in order to verify the results from the modal analysis, in which a validated finite element model was obtained via a model-updating method that considers changes in the in situ experimental data. By using a validated finite element model, the equivalent static load under SSE conditions stipulated by the KEPIC MFA gave a stress of 135 MPa that occurred at the connections of the stem and body. A larger stress of 183 MPa was induced when we used a CQC method with a design response spectrum that uses 2% damping ratio. These values were lower than the allowable strength of the materials used for manufacturing the butterfly valve, and, therefore, its structural safety met the KEPIC MFA requirements. PMID:24955416

  20. Conditions for the successful integration of Human and Organizational Factors (HOF) in the nuclear safety analysis.

    PubMed

    Tosello, Michèle; Lévêque, Françoise; Dutillieu, Stéphanie; Hernandez, Guillaume; Vautier, Jean-François

    2012-01-01

    This communication presents some elements which come from the experience feedback at CEA about the conditions for the successful integration of HOF in the nuclear safety analysis. To point out some of these conditions, one of the concepts proposed by Edgar Morin to describe the functioning of "complex" systems: the dialogical principle has been used. The idea is to look for some dialogical pairs. The elements of this kind of pair are both complementary and antagonist to one another. Three dialogical pairs are presented in this communication. The first two pairs are related to the organization of the HOF network and the last one is related to the methods which are used to analyse the working situations. The three pairs are: specialist - non-specialist actors of the network, centralized - distributed human resources in the network and microscopic - macroscopic levels of HOF methods to analyse the working situations. To continuously improve these three dialogical pairs, it is important to keep the differences which exist between the two elements of a pair and to find and maintain a balance between the two elements of the pairs.