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Sample records for nuclear safety center

  1. 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)

  2. 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,…

  3. 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

  4. 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…

  5. 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.

  6. 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…

  7. 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.

  8. 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.

  9. 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").

  10. 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…

  11. 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.

  12. 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

  13. 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.

  14. 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.

  15. 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.

  16. Patient Safety Center Organization

    DTIC Science & Technology

    2006-06-01

    Enterectomy Bariatric 7/20/05 4 hours/course (28 hours/year) Surgery R 7 Dr. Karen Horvath R5 Lap Enterectomy & Colectomy 11/30/05 4...areas in the UW Schools of Nursing and Dentistry, at the Harborview Research Center Microvascular Surgery lab, with the Seattle Children’s Hospital and...1 laboratory complex (2500 sq ft) has been designed within the University of Washington Medical Center, in the Surgery Pavilion complex

  17. 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…

  18. 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.

  19. 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.

  20. 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.

  1. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Nuclear safety. 923.7001... Efficiency, Renewable Energy Technologies, and Occupational Safety Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under its own statutory authority derived from...

  2. 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}.

  3. 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.

  4. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 48 Federal Acquisition Regulations System 5 2012-10-01 2012-10-01 false Nuclear safety. 923.7001 Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS... Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under...

  5. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 48 Federal Acquisition Regulations System 5 2014-10-01 2014-10-01 false Nuclear safety. 923.7001 Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS... Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under...

  6. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 48 Federal Acquisition Regulations System 5 2013-10-01 2013-10-01 false Nuclear safety. 923.7001 Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS... Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under...

  7. 48 CFR 923.7001 - Nuclear safety.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 48 Federal Acquisition Regulations System 5 2011-10-01 2011-10-01 false Nuclear safety. 923.7001 Section 923.7001 Federal Acquisition Regulations System DEPARTMENT OF ENERGY SOCIOECONOMIC PROGRAMS... Programs 923.7001 Nuclear safety. The DOE regulates the nuclear safety of its major facilities under...

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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

  14. 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.

  15. 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…

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. 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…

  3. 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.

  4. 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.

  5. 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.

  6. Nuclear Wallet Cards from the National Nuclear Data Center (NNDC)

    DOE Data Explorer

    Tuli, Jagdish K.

    Nuclear Wallet Cards present properties for ground and isomeric states of all known nuclides. Properties given are: spin and parity assignments, nuclear mass excesses, half-life, isotopic abundances, and decay modes. Appendices contain properties of elements, fundamental constants and other useful information. Nuclear Wallet Cards booklet is published by the National Nuclear Data Center and its electronic (current) version is periodically updated. The Nuclear Wallet Cards by Dr. Jagdish K. Tuli, presently in its 8th edition, is distributed in print as well as in PDA-adaptable Palm Pilot format; the data table as an ASCII file is available upon request. [Taken from http://www.nndc.bnl.gov/wallet/

  7. 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.

  8. 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.

  9. 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.

  10. National Center for Nuclear Security - NCNS

    SciTech Connect

    2014-11-12

    As the United States embarks on a new era of nuclear arms control, the tools for treaty verification must be accurate and reliable, and must work at stand-off distances. The National Center for Nuclear Security, or NCNS, at the Nevada National Security Site, is poised to become the proving ground for these technologies. The center is a unique test bed for non-proliferation and arms control treaty verification technologies. The NNSS is an ideal location for these kinds of activities because of its multiple environments; its cadre of experienced nuclear personnel, and the artifacts of atmospheric and underground nuclear weapons explosions. The NCNS will provide future treaty negotiators with solid data on verification and inspection regimes and a realistic environment in which future treaty verification specialists can be trained. Work on warhead monitoring at the NCNS will also support future arms reduction treaties.

  11. National Center for Nuclear Security - NCNS

    ScienceCinema

    None

    2016-07-12

    As the United States embarks on a new era of nuclear arms control, the tools for treaty verification must be accurate and reliable, and must work at stand-off distances. The National Center for Nuclear Security, or NCNS, at the Nevada National Security Site, is poised to become the proving ground for these technologies. The center is a unique test bed for non-proliferation and arms control treaty verification technologies. The NNSS is an ideal location for these kinds of activities because of its multiple environments; its cadre of experienced nuclear personnel, and the artifacts of atmospheric and underground nuclear weapons explosions. The NCNS will provide future treaty negotiators with solid data on verification and inspection regimes and a realistic environment in which future treaty verification specialists can be trained. Work on warhead monitoring at the NCNS will also support future arms reduction treaties.

  12. Current Development of Nuclear Thermal Propulsion technologies at the Center for Space Nuclear Research

    SciTech Connect

    Robert C. O'Brien; Steven K. Cook; Nathan D. Jerred; Steven D. Howe; Ronald Samborsky; Daniel Brasuell

    2012-09-01

    Nuclear power and propulsion has been considered for space applications since the 1950s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors / rocket engines in the Rover/NERVA programs1. The Aerojet Corporation was the prime contractor for the NERVA program. Modern changes in environmental laws present challenges for the redevelopment of the nuclear rocket. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel composition that is significantly different from those of the NERVA project can be engineered; this may be needed to ensure public support and compliance with safety requirements. The Center for Space Nuclear Research (CSNR) is pursuing a number of technologies, modeling and testing processes to further the development of safe, practical and affordable nuclear thermal propulsion systems.

  13. 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.

  14. 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.

  15. 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).

  16. 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.

  17. 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.

  18. NUCLEAR INFORMATION SERVICES AT THE NATIONAL NUCLEAR DATA CENTER.

    SciTech Connect

    BURROWS,T.W.; DUNFORD,C.L.

    2004-09-26

    The National Nuclear Data Center has provided remote access to its databases and other resources since 1986. This year we have completed the modernization of our databases and Web site. Resources available from our Web site will be summarized and some of the major improvements described in more detail.

  19. 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.

  20. Colleges and Universities Highway Traffic and Safety Centers.

    ERIC Educational Resources Information Center

    Aaron, James E., Ed.; Ritzel, Dale O., Ed.

    After consideration of the organizing of university safety centers and the growth and role of such centers in the future, descriptions are presented of the activities and practices in each of 16 existing college and university highway traffic and safety centers. Information is presented regarding center objectives, programs, staff composition,…

  1. 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.

  2. 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)

  3. 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.

  4. 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.

  5. 46 CFR 50.10-23 - Marine Safety Center.

    Code of Federal Regulations, 2010 CFR

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Marine Safety Center. 50.10-23 Section 50.10-23 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS Definition of Terms Used in This Subchapter § 50.10-23 Marine Safety Center. The term Marine Safety...

  6. 46 CFR 50.10-23 - Marine Safety Center.

    Code of Federal Regulations, 2014 CFR

    2014-10-01

    ... 46 Shipping 2 2014-10-01 2014-10-01 false Marine Safety Center. 50.10-23 Section 50.10-23 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS Definition of Terms Used in This Subchapter § 50.10-23 Marine Safety Center. The term Marine Safety...

  7. 46 CFR 50.10-23 - Marine Safety Center.

    Code of Federal Regulations, 2013 CFR

    2013-10-01

    ... 46 Shipping 2 2013-10-01 2013-10-01 false Marine Safety Center. 50.10-23 Section 50.10-23 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS Definition of Terms Used in This Subchapter § 50.10-23 Marine Safety Center. The term Marine Safety...

  8. 46 CFR 50.10-23 - Marine Safety Center.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 2 2011-10-01 2011-10-01 false Marine Safety Center. 50.10-23 Section 50.10-23 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS Definition of Terms Used in This Subchapter § 50.10-23 Marine Safety Center. The term Marine Safety...

  9. 46 CFR 50.10-23 - Marine Safety Center.

    Code of Federal Regulations, 2012 CFR

    2012-10-01

    ... 46 Shipping 2 2012-10-01 2012-10-01 false Marine Safety Center. 50.10-23 Section 50.10-23 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING GENERAL PROVISIONS Definition of Terms Used in This Subchapter § 50.10-23 Marine Safety Center. The term Marine Safety...

  10. 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.

  11. 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.

  12. 76 FR 39811 - International Center for Technology Assessment and the Center for Food Safety; Noxious Weed...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-07

    ... International Center for Technology Assessment and the Center for Food Safety (the petitioners) requested that...; ] DEPARTMENT OF AGRICULTURE Animal and Plant Health Inspection Service International Center for Technology Assessment and the Center for Food Safety; Noxious Weed Status of Kentucky Bluegrass Genetically...

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  1. 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 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  2. 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 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  3. 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 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  4. 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 72.124 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS...

  5. Nuclear energy center finance and ownership considerations

    SciTech Connect

    Morris, J.A.; Wilder, R.P.

    1980-09-01

    Finance and ownership alternatives for a nuclear energy center (NEC) in South Carolina are analyzed in the context of the capital market and tax differences among alternatives. The ownership alternatives considered are (1) the private or private/public joint venture, (2) full public ownership and (3) a hybrid ownership form featuring federal involvement in the initial site development and permit phase, followed by a transition to private ownership. Public ownership is associated with considerably lower out-of-pocket costs than private ownership; the difference between the two, however, is related to subsidies from other parts of society to electricity customers of a publicly owned NEC. The attitudes of participating utilities on ownership forms are examined, with the finding of general strong opposition to increased federal involvement in the electric utility industry through NEC ownership. The conclusion is that the private-private/public joint venture is the preferable ownership form and that public ownership should be employed only if the private sector fails to respond to future energy demand.

  6. 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…

  7. 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.

  8. Stennis Space Center observes 2009 Safety and Health Day

    NASA Technical Reports Server (NTRS)

    2009-01-01

    Sue Smith, a medical clinic employee at NASA's John C. Stennis Space Center, takes the temperature of colleague Karen Badon during 2009 Safety and Health Day activities Oct. 22. Safety Day activities included speakers, informational sessions and a number of displays on safety and health issues. Astronaut Dominic Gorie also visited the south Mississippi rocket engine testing facility during the day to address employees and present several Silver Snoopy awards for outstanding contributions to flight safety and mission success. The activities were part of an ongoing safety and health emphasis at Stennis.

  9. 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…

  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. 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.

  12. Introducing Safety Topics Using a Student-Centered Approach

    ERIC Educational Resources Information Center

    Wright, Steven M.

    2005-01-01

    The activities that introduce topics of chemical health and safety are offered, using student-centered cooperative method. Teaching methods and exercises used in various activities were considered valuable enough to be modified and published in a book of activities for safety.

  13. 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.

  14. 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.

  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. 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…

  17. 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.

  18. National Center for Nuclear Security: The Nuclear Forensics Project (F2012)

    SciTech Connect

    Klingensmith, A. L.

    2012-03-21

    These presentation visuals introduce the National Center for Nuclear Security. Its chartered mission is to enhance the Nation’s verification and detection capabilities in support of nuclear arms control and nonproliferation through R&D activities at the NNSS. It has three focus areas: Treaty Verification Technologies, Nonproliferation Technologies, and Technical Nuclear Forensics. The objectives of nuclear forensics are to reduce uncertainty in the nuclear forensics process & improve the scientific defensibility of nuclear forensics conclusions when applied to nearsurface nuclear detonations. Research is in four key areas: Nuclear Physics, Debris collection and analysis, Prompt diagnostics, and Radiochemistry.

  19. 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.

  20. 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.

  1. 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.

  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. 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.

  4. 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

    ... Response to Recommendation 2011-1 of the Defense Nuclear Facilities Safety Board, Safety Culture at the..., concerning Safety Culture at the Waste Treatment and Immobilization Plant, to the Department of Energy. In...) acknowledges receipt of Defense Nuclear Facilities Safety Board (Board) Recommendation 2011-1, Safety...

  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. Chart of Nuclides from the National Nuclear Data Center (NNDC)

    DOE Data Explorer

    The Chart of Nuclides is a software product that allows users to search and plot nuclear structure and nuclear decay data interactively. The Chart of Nuclides was developed by the National Nuclear Data Center (NNDC). It provides an interface between web users and several databases containing nuclear structure, nuclear decay and some neutron-induced nuclear reaction information. Using the Chart of Nuclides, it is possible to search for nuclear level properties (energy, half-life, spin-parity), gamma-ray information (energy, intensity, multipolarity, coincidences),radiation information following nuclear decay (energy, intensity, dose), and neutron-induced reaction data from the BNL-325 book (thermal cross section and resonance integral). The information provided by the Chart of Nuclides can be seen in tables, level schemes and an interactive chart of nuclides. (From the Chart of Nuclides Description at http://www.nndc.bnl.gov/chart/help/index.jsp?product=chart)

  7. 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.

  8. 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.

  9. 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

  10. Texas School Safety Center: Charting the Course for School Safety in Turbulent Times.

    ERIC Educational Resources Information Center

    Renick, Judy

    2002-01-01

    Describes the Texas School Safety Center, established by the Texas legislature to provide interdisciplinary resources to promote school safety statewide through activities such as regional training institutes, youth leadership training, and technical assistance. Provides an example from the "Proactive Guide for Threat of Terrorism in Schools,"…

  11. Patient safety goals for the proposed Federal Health Information Technology Safety Center.

    PubMed

    Sittig, Dean F; Classen, David C; Singh, Hardeep

    2015-03-01

    The Office of the National Coordinator for Health Information Technology is expected to oversee creation of a Health Information Technology (HIT) Safety Center. While its functions are still being defined, the center is envisioned as a public-private entity focusing on promotion of HIT related patient safety. We propose that the HIT Safety Center leverages its unique position to work with key administrative and policy stakeholders, healthcare organizations (HCOs), and HIT vendors to achieve four goals: (1) facilitate creation of a nationwide 'post-marketing' surveillance system to monitor HIT related safety events; (2) develop methods and governance structures to support investigation of major HIT related safety events; (3) create the infrastructure and methods needed to carry out random assessments of HIT related safety in complex HCOs; and (4) advocate for HIT safety with government and private entities. The convening ability of a federally supported HIT Safety Center could be critically important to our transformation to a safe and effective HIT enabled healthcare system.

  12. Southwest Border Food Safety and Defense Center: Creative Ideas for Promoting Food Safety and Food Protection

    ERIC Educational Resources Information Center

    Koukel, Sonja

    2015-01-01

    Foodborne illness has a significant impact on public health and consumer confidence in the U.S. The Southwest Border Food Safety and Defense Center was established to provide educational programs, trainings, and workshops to address the health and well-being of consumers as it relates to food safety and food protection. A partnership between New…

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center

    SciTech Connect

    Myers, Astasia

    2011-06-28

    The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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).

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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).

  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. 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.

  14. 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.

  15. 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)

  16. 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.

  17. 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.

  18. 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.

  19. Nudat: Nuclear Structure and Decay Data from the National Nuclear Data Center (NNDC)

    DOE Data Explorer

    NuDat allows users to search and plot nuclear structure and decay data interactively. NuDat was developed by the National Nuclear Data Center (NNDC)but utilizes contributions from physicists around the world. It provides an interface between web users and several databases containing nuclear structure, nuclear decay and some neutron-induced nuclear reaction information. Users can search for nuclear level properties (energy, half-life, spinparity), gamma-ray information (energy, intensity, multipolarity, coincidences), radiation information following nuclear decay (energy, intensity, dose), and neutron-induced reaction data from the BNL-325 book (thermal cross section and resonance integral). The information provided by NuDat 2 can be viewed in tables, level schemes and an interactive chart of nuclides.

  20. Shaping NASA's Kennedy Space Center Safety for the Future

    NASA Technical Reports Server (NTRS)

    Kirkpatrick, Paul; McDaniel, Laura; Smith, Maynette

    2011-01-01

    With the completion of the Space Shuttle Program, the Kennedy Space Center (KSC) safety function will be required to evolve beyond the single launch vehicle launch site focus that has held prominence for almost fifty years. This paper will discuss how that evolution is taking place. Specifically, we will discuss the future of safety as it relates to a site that will have multiple, very disparate, functions. These functions will include new business; KSC facilities not under the control of NASA; traditional payload and launch vehicle processing; and, operations conducted by NASA personnel, NASA contractors or a combination of both. A key element in this process is the adaptation of the current KSC set of safety requirements into a multi-faceted set that can address each of the functions above, while maintaining our world class safety environment. One of the biggest challenges that will be addressed is how to protect our personnel and property without dictating how other Non-NASA organizations protect their own employees and property. The past history of KSC Safety will be described and how the lessons learned from previous programs will be applied to the future. The lessons learned from this process will also be discussed as information for other locations that may undergo such a transformation.

  1. 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...

  2. Los Alamos Neutron Science Center (LANSCE) Nuclear Science Facilities

    SciTech Connect

    Nelson, Ronald Owen; Wender, Steve

    2015-06-19

    The Los Alamos Neutron Science Center (LANSCE) facilities for Nuclear Science consist of a high-energy "white" neutron source (Target 4) with 6 flight paths, three low-energy nuclear science flight paths at the Lujan Center, and a proton reaction area. The neutron beams produced at the Target 4 complement those produced at the Lujan Center because they are of much higher energy and have shorter pulse widths. The neutron sources are driven by the 800-MeV proton beam of the LANSCE linear accelerator. With these facilities, LANSCE is able to deliver neutrons with energies ranging from a milli-electron volt to several hundreds of MeV, as well as proton beams with a wide range of energy, time and intensity characteristics. The facilities, instruments and research programs are described briefly.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. NUCLEAR FORENSICS ANALYSIS CENTER FORENSIC ANALYSIS TO DATA INTERPRETATION

    SciTech Connect

    Nichols, T.

    2011-02-07

    The Nuclear Forensics Analysis Center (NFAC) is part of Savannah River National Laboratory (SRNL) and is one of only two USG National Laboratories accredited to perform nuclear forensic analyses to the requirements of ISO 17025. SRNL NFAC is capable of analyzing nuclear and radiological samples from bulk material to ultra-trace samples. NFAC provides analytical support to the FBI's Radiological Evidence Examination Facility (REEF), which is located within SRNL. REEF gives the FBI the capability to perform traditional forensics on material that is radiological and/or is contaminated. SRNL is engaged in research and development efforts to improve the USG technical nuclear forensics capabilities. Research includes improving predictive signatures and developing a database containing comparative samples.

  8. 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.

  9. 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.

  10. 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.

  11. Safety Evaluation Report for the Claiborne Enrichment Center, Homer, Louisiana (Docket No. 70-3070)

    SciTech Connect

    Not Available

    1994-01-01

    This report documents the US Nuclear Regulatory Commission (NRC) staff review and safety evaluation of the Louisiana Energy Services, L.P. (LES, the applicant) application for a license to possess and use byproduct, source, and special nuclear material and to enrich natural uranium to a maximum of 5 percent U-235 by the gas centrifuge process. The plant, to be known as the Claiborne Enrichment Center (CEC), would be constructed near the town of Homer in Claiborne Parish, Louisiana. At full production in a given year, the plant will receive approximately 4,700 tonnes of feed UF{sub 6} and produce 870 tonnes of low-enriched UF{sub 6}, and 3,830 tonnes of depleted UF{sub 6} tails. Facility construction, operation, and decommissioning are expected to last 5, 30, and 7 years, respectively. The objective of the review is to evaluate the potential adverse impacts of operation of the facility on worker and public health and safety under both normal operating and accident conditions. The review also considers the management organization, administrative programs, and financial qualifications provided to assure safe design and operation of the facility. The NRC staff concludes that the applicant`s descriptions, specifications, and analyses provide an adequate basis for safety review of facility operations and that construction and operation of the facility does not pose an undue risk to public health and safety.

  12. RECENT ACTIVITIES AT THE CENTER FOR SPACE NUCLEAR RESEARCH FOR DEVELOPING NUCLEAR THERMAL ROCKETS

    SciTech Connect

    Robert C. O'Brien

    2001-09-01

    Nuclear power has been considered for space applications since the 1960s. Between 1955 and 1972 the US built and tested over twenty nuclear reactors/ rocket-engines in the Rover/NERVA programs. However, changes in environmental laws may make the redevelopment of the nuclear rocket more difficult. Recent advances in fuel fabrication and testing options indicate that a nuclear rocket with a fuel form significantly different from NERVA may be needed to ensure public support. The Center for Space Nuclear Research (CSNR) is pursuing development of tungsten based fuels for use in a NTR, for a surface power reactor, and to encapsulate radioisotope power sources. The CSNR Summer Fellows program has investigated the feasibility of several missions enabled by the NTR. The potential mission benefits of a nuclear rocket, historical achievements of the previous programs, and recent investigations into alternatives in design and materials for future systems will be discussed.

  13. 77 FR 7139 - Public Availability of Defense Nuclear Facilities Safety Board; FY 2010 Service Contract...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-02-10

    ... From the Federal Register Online via the Government Publishing Office DEFENSE NUCLEAR FACILITIES SAFETY BOARD Public Availability of Defense Nuclear Facilities Safety Board; FY 2010 Service Contract Inventory Analysis/FY 2011 Service Contract Inventory AGENCY: Defense Nuclear Facilities Safety Board...

  14. 78 FR 12042 - Public Availability of Defense Nuclear Facilities Safety Board FY 2011 Service Contract Inventory...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-21

    ... From the Federal Register Online via the Government Publishing Office DEFENSE NUCLEAR FACILITIES SAFETY BOARD Public Availability of Defense Nuclear Facilities Safety Board FY 2011 Service Contract Inventory Analysis/FY 2012 Service Contract Inventory AGENCY: Defense Nuclear Facilities Safety Board...

  15. Transfusion safety: is this the business of blood centers?

    PubMed

    Slapak, Colleen; Fredrich, Nanci; Wagner, Jeffrey

    2011-12-01

    ATSO is in a unique position to break down organizational silos between hospitals and blood centers through the development of a collaborative relationship between the two entities. Use of the TSO as blood center staff centralizes the role into a consultative position thereby retaining the independence of the hospitals. The TSO position then becomes a value-added service offered by the blood center designed to supplement processes within the hospital.Whether the TSO is based in the hospital or the blood center, improvements are gained through appropriate utilization of blood components, reductions in hospital costs, ongoing education of hospital staff involved in transfusion practice, and increased availability of blood products within the community. Implementation and standardization of best practice processes for ordering and administration of blood products developed by TSOs leads to improved patient outcomes. As a liaison between hospitals and blood centers, the TSO integrates the mutual goal of transfusion safety: the provision of the safest blood product to the right patient at the right time for the right reason.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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,...

  1. 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 ...

  2. Westar's Lawrence Energy Center wins for not blinking on safety

    SciTech Connect

    Peltier, R.

    2007-07-15

    It took Westar Energy eight years to upgrade the Lawrence Energy Center to burn Powder River Basin coal. Its zero lost-time accident record during the eight-year, million-man-hour project is a testament to Westar's commitment to workplace safety. The plant won the Powder River Basin Coal Users' Group plant of the year award for 2006. The article describes all the changes implemented at the plant, including replacing and upgrading controls for the belt conveyor, replacing the coal crushers, minimising dust and modifying coal bunkers, to cope with the increased volatility of Powder River Basin coal. Modifications were made to minimise slagging and fouling of boilers. 10 photos.

  3. 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.

  4. Equipment for nuclear medical centers, production capabilities of Rosatom enterprises

    NASA Astrophysics Data System (ADS)

    Gavrish, Yu. N.; Koloskov, S. A.; Smirnov, V. P.; Strokach, A. P.

    2015-12-01

    Analysis of the capabilities of the State Corporation Rosatom enterprises on the development and production of diagnostic and therapeutic equipment for nuclear medicine centers is presented. Prospects of the development of accelerator equipment for the production of a wide range of radioisotope products are shown, and the trends of its development are determined. A comparative analysis of the technical parameters of domestic tomographs and devices for brachytherapy with foreign counterparts is given.

  5. Equipment for nuclear medical centers, production capabilities of Rosatom enterprises

    SciTech Connect

    Gavrish, Yu. N.; Koloskov, S. A.; Smirnov, V. P.; Strokach, A. P.

    2015-12-15

    Analysis of the capabilities of the State Corporation Rosatom enterprises on the development and production of diagnostic and therapeutic equipment for nuclear medicine centers is presented. Prospects of the development of accelerator equipment for the production of a wide range of radioisotope products are shown, and the trends of its development are determined. A comparative analysis of the technical parameters of domestic tomographs and devices for brachytherapy with foreign counterparts is given.

  6. Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

    SciTech Connect

    Todd R. Allen

    2011-12-01

    This is a document required by Basic Energy Sciences as part of a mid-term review, in the third year of the five-year award period and is intended to provide a critical assessment of the Center for Materials Science of Nuclear Fuels (strategic vision, scientific plans and progress, and technical accomplishments).

  7. 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.

  8. 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.

  9. Educators benefit from energy information centers at nuclear plant sites

    SciTech Connect

    Krcma-Olson, L.

    1994-12-31

    While issues like dry storage, low-level waste storage, radiation, and license extension are projects with a technical perspective that need to be planned and executed at nuclear power plants, more difficult is the political perspective-gaining public acceptance to allow these projects to proceed. And public perception is predicated on the way plant neighbors and community members understand, accept, and trust the plants. Community educators are a key audience. Annually, U.S. information centers host about one million visitors; roughly half of them are school children who will soon join the ranks of voters, taxpayers, utility customers, and employees. Programs for educators and their classes vary from tours of centers that include computer games and video programs on energy-related topics to audio-visual presentations by center personnel. Some facilities have environmental activities such as hatcheries or nature trails, while others offer plant tours to specific age groups.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. Electric power transmission for a Hanford Nuclear Energy Center (HNEC)

    SciTech Connect

    Harty, H.; Dowis, W.J.

    1983-06-01

    The original study of transmission for a Hanford Nuclear Energy Center (HNEC), which was completed in September 1975, was updated in June 1978. The present 1983 revision takes cognizance of recent changes in the electric power situation of the PNW with respect to: (1) forecasts of load growth, (2) the feasibility of early use of 1100 kV transmission, and (3) the narrowing opportunities for siting nuclear plants in the region. The purpose of this update is to explore and describe additions to the existing transmission system that would be necessary to accommodate three levels of generation at HNEC. Comparisons with a PNW system having new thermal generating capacity distributed throughout the marketing region are not made as was done in earlier versions.

  18. 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.

  19. 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.

  20. 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.

  1. 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

  2. NASA Engineering and Safety Center NDE Super Problem Resolution Team

    NASA Astrophysics Data System (ADS)

    Prosser, W. H.

    2007-03-01

    The NASA Engineering and Safety Center (NESC) is an independent organization, which was charted in the wake of the Space Shuttle Columbia accident to serve as an Agency-wide technical resource focused on engineering excellence. The objective of the NESC is to improve safety by performing in-depth independent engineering assessments, testing, and analysis to uncover technical vulnerabilities and to determine appropriate preventative and corrective actions for problems, trends or issues within NASA's programs, projects and institutions. Critical to the NESC are teams of experts in a number of core disciplines including nondestructive evaluation (NDE). These teams, designated Super Problem Resolution Teams (SPRTs), draw upon the best engineering expertise from across the Agency and include partnerships with other government agencies, national laboratories, universities and industry. The NESC NDE SPRT provides a ready resource of NDE technical expertise to support NESC Independent Technical Assessments and Investigations. The purpose of this session will be to provide an overview of the NESC and the NDE SPRT along with a few examples of NDE related problems that the team has addressed for NASA Programs. It is hoped that this session will be of interest to the general NDE community and will foster contacts with additional NDE experts that might provide future support to the NASA NESC NDE SPRT.

  3. 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...

  4. 77 FR 30029 - Entergy Nuclear Operations, Inc.; Establishment of Atomic Safety and Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-05-21

    ... COMMISSION Entergy Nuclear Operations, Inc.; Establishment of Atomic Safety and Licensing Board Pursuant to..., notice is hereby given that an Atomic Safety and Licensing Board (Board) is being established to preside..., Atomic Safety and Licensing Board Panel, U.S. Nuclear Regulatory Commission, Washington, DC...

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 76 FR 40733 - National Institute for Occupational Safety and Health, (NIOSH), World Trade Center Health Program...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-11

    ... From the Federal Register Online via the Government Publishing Office DEPARTMENT OF HEALTH AND HUMAN SERVICES Centers for Disease Control and Prevention National Institute for Occupational Safety and Health, (NIOSH), World Trade Center Health Program Science/Technical Advisory Committee...

  12. 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.

  13. SAID Partial Wave Analyses from CNS DAC (Center for Nuclear Studies Data Analysis Center)

    DOE Data Explorer

    George Washington University (GW) has one of the largest university-based nuclear-physics groups in the nation. Many of the current and future projects are geared to Thomas Jefferson National Accelerator Facility (JLab) at Newport News, VA. JLab is the world's premier electron accelerator for nuclear physics, and GW is one of the charter members of the governing body of JLab, the Southeastern Universities Research Association (SURA). The George Washington Data Analysis Center (DAC) was created in 1998 by an agreement among the Department of Energy, Jefferson Lab, and the GW Center for Nuclear Studies.The activities of the DAC fall into four distinct categories: 1) Performing partial-wave analyses of fundamental two- and three-body reactions; 2) Maintenance of databases associated with these reactions; 3) Development of software to disseminate DAC results (as well as the results of competing model-independent analyses and potential approaches); and 4) Phenomenological and theoretical investigations which bridge the gap between theory and experiment; in particular, the extraction of N* and D * hadronic and electromagnetic couplings. Partial Wave Analyses (and the associated databases) available at GW are: Pion-Nucleon, Kaon-Nucleon, Nucleon-Nucleon, Pion Photoproduction, Pion Electroproduction, Kaon Photoproduction, Eta Photoproduction, Eta-Prime Photoproduction, Pion-Deuteron (elastic), and Pion-Deuteron to Proton+Proton. [Taken from http://www.gwu.edu/~ndl/dac.htm">http://www.gwu.edu/~ndl/dac.htm

  14. 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...

  15. 10 CFR 1.42 - Office of Nuclear Material Safety and Safeguards.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... 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 2011-01-01 2011-01-01 false Office of Nuclear Material Safety and Safeguards....

  16. 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...

  17. 10 CFR 1.42 - Office of Nuclear Material Safety and Safeguards.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... 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 2010-01-01 2010-01-01 false Office of Nuclear Material Safety and Safeguards....

  18. 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...

  19. 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...

  20. 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...

  1. 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...

  2. 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...

  3. 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.

  4. 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.

  5. Recent Developments in Nuclear Data Measurement capabilities at the Gaerttner LINAC Center at RPI

    NASA Astrophysics Data System (ADS)

    Danon, Y.; Daskalakis, A.; McDermott, B.; Thompson, N.; Youmans, A.; Block, R.; Barry, D.; Epping, B.; Leinweber, G.; Rapp, M.; Donovan, T.

    2016-03-01

    The Gaerttner LINAC Center at RPI uses a 60 MeV electron linear accelerator to produce short pulses of neutrons with duration of 5-5000 ns. The main research thrust at the Center is nuclear data for nuclear reactors and criticality safety applications. The Center includes several setups for time-of-flight measurements including neutron transmission, capture and scattering detectors, and a lead slowing-down spectrometer. Experiments were designed to produce neutron interaction cross sections that cover the energy range of 0.01 eV to 20 MeV. Recently added experiments include: setups for keV and fast neutron transmission, a C6D6 detector array for keV neutron capture measurements, and a fast neutron scattering system. Results discussed here include fast neutron scattering and angular distributions for natFe, iron capture measurements for incident neutrons from 1 keV to 2 MeV, fast neutron transmission through W and H2O samples, and keV transmission through Mo isotopes.

  6. Feed Materials Production Center environmental, safety and health management plan: (Revision to NLCO-2037 preliminary)

    SciTech Connect

    Loudin, D.J.

    1986-09-01

    The Feed Materials Production Center (FMPC) produces uranium metal for DOE defense programs at the Washington and South Carolina reactor sites, and DOE facilities elsewhere. Since the FMPC produces uranium metal products in support of various DOE defense programs, it is important that the FMPC maintain high standards of operation in a safe and environmentally compatible manner. This Environmental, Safety and Health (ES and H) Management Plan is a vital initial step to bring together all of the ES and H programs as an integrated plan which meets site ES and H concerns, and supports continued operations to fulfill FMPC mission requirements. The ES and H Management Plan report explains the FMPC mission and history, describes the site and surrounding area, and details the purpose and organization of the report; describes Air Pollution Control, Water Pollution Control, Solid Waste Management, and the Remedial Action Plan at the FMPC; describes the Safety Analyses program, Health Physics/Radiation Protection, Nuclear Criticality Prevention, Industrial Hygiene, and Safety/Fire Protection programs; and describes the FMPC Emergency Preparedness Plan.

  7. Energy Frontier Research Center, Center for Materials Science of Nuclear Fuels

    SciTech Connect

    Todd R. Allen, Director

    2011-04-01

    The Office of Science, Basic Energy Sciences, has funded the INL as one of the Energy Frontier Research Centers in the area of material science of nuclear fuels. This document is the required annual report to the Office of Science that outlines the accomplishments for the period of May 2010 through April 2011. The aim of the Center for Material Science of Nuclear Fuels (CMSNF) is to establish the foundation for predictive understanding of the effects of irradiation-induced defects on thermal transport in oxide nuclear fuels. The science driver of the center’s investigation is to understand how complex defect and microstructures affect phonon mediated thermal transport in UO2, and achieve this understanding for the particular case of irradiation-induced defects and microstructures. The center’s research thus includes modeling and measurement of thermal transport in oxide fuels with different levels of impurities, lattice disorder and irradiation-induced microstructure, as well as theoretical and experimental investigation of the evolution of disorder, stoichiometry and microstructure in nuclear fuel under irradiation. With the premise that thermal transport in irradiated UO2 is a phonon-mediated energy transport process in a crystalline material with defects and microstructure, a step-by-step approach will be utilized to understand the effects of types of defects and microstructures on the collective phonon dynamics in irradiated UO2. Our efforts under the thermal transport thrust involved both measurement of diffusive phonon transport (an approach that integrates over the entire phonon spectrum) and spectroscopic measurements of phonon attenuation/lifetime and phonon dispersion. Our distinct experimental efforts dovetail with our modeling effort involving atomistic simulation of phonon transport and prediction of lattice thermal conductivity using the Boltzmann transport framework.

  8. 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).

  9. 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.

  10. Progress of Covariance Evaluation at the China Nuclear Data Center

    SciTech Connect

    Xu, R.; Zhang, Q.; Zhang, Y.; Liu, T.; Ge, Z.; Lu, H.; Sun, Z.; Yu, B.; Tang, G.

    2015-01-15

    Covariance evaluations at the China Nuclear Data Center focus on the cross sections of structural materials and actinides in the fast neutron energy range. In addition to the well-known Least-squares approach, a method based on the analysis of the sources of experimental uncertainties is especially introduced to generate a covariance matrix for a particular reaction for which multiple measurements are available. The scheme of the covariance evaluation flow is presented, and an example of n+{sup 90}Zr is given to illuminate the whole procedure. It is proven that the accuracy of measurements can be properly incorporated into the covariance and the long-standing small uncertainty problem can be avoided.

  11. Center for Nuclear Medicine Research in Alzheimer`s Disease Health Sciences Center, West Virginia University. Environmental Assessment

    SciTech Connect

    Not Available

    1994-04-01

    The Environmental Assessment (EA) of the Center for Nuclear Medicine Research in Alzheimer`s Disease (CNMR) at the Health Sciences Center, at West Virginia University in Morgantown, West Virginia for the construction and operation was prepared by DOE. The EA documents analysis of the environmental and socioeconomic impacts that might occur as a result of these actions, and characterizes potential impacts on the environment. In the EA, DOE presents its evaluation of potential impacts of construction and operation of the CNMR on health and safety of both workers and the public, as well as on the external environment. Construction impacts include the effects of erosion, waste disposal, air emissions, noise, and construction traffic and parking. Operational impacts include the effects of waste generation (domestic, sanitary, hazardous, medical/biological, radioactive and mixed wastes), radiation exposures, air emissions (radioactive, criteria, and air toxics), noise, and new workers. No sensitive resources (wetlands, special sources of groundwater, protected species) exist in the area of project effect.

  12. 78 FR 37228 - Cooperative Agreement To Support the Western Center for Food Safety

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-20

    ... HUMAN SERVICES Food and Drug Administration Cooperative Agreement To Support the Western Center for Food Safety AGENCY: Food and Drug Administration, HHS. ACTION: Notice. SUMMARY: The Food and Drug... Western Center for Food Safety (WCFS). FDA regards the continued support of WCFS as crucial to...

  13. 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.

  14. 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.

  15. 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.

  16. 46 CFR 1.03-30 - Appeals from decisions or actions of the Marine Safety Center.

    Code of Federal Regulations, 2011 CFR

    2011-10-01

    ... 46 Shipping 1 2011-10-01 2011-10-01 false Appeals from decisions or actions of the Marine Safety... APPLICABLE TO THE PUBLIC ORGANIZATION, GENERAL COURSE AND METHODS GOVERNING MARINE SAFETY FUNCTIONS Rights of Appeal § 1.03-30 Appeals from decisions or actions of the Marine Safety Center. (a) Any person...

  17. Center Director Bridges opens Super Safety and Health Day at KSC.

    NASA Technical Reports Server (NTRS)

    1999-01-01

    Center Director Roy Bridges opens the second Super Safety and Health Day at Kennedy Space Center, an entire day when most normal work activities are suspended to allow personnel to attend safety- and health-related activities. The theme, 'Safety and Health Go Hand in Hand,' emphasized KSC's commitment to place the safety and health of the public, astronauts, employees and space- related resources first and foremost. Events included a keynote address, a panel session about related issues, vendor exhibits, and safety training in work groups. The keynote address and panel session were also broadcast internally over NASA television.

  18. Evaluated Nuclear Structure Data File (ENSDF) from the National Nuclear Data Center (NNDC)

    DOE Data Explorer

    ENSDF contains evaluated nuclear structure and decay data in a standard format. An international network of evaluators contributes to the database, which is maintained by the National Nuclear Data Center at Brookhaven National Laboratory. Information in the database is regularly updated to reflect revised evaluation results. Most of the recently completed evaluations are published in Nuclear Data Sheets, a monthly journal published by Academic Press, a division of Elsevier Science. For each nuclide, all known experimental data used to deduce nuclear structure information are included. Each type of experiment is presented as a separate dataset. In addition, there is a dataset of "adopted" level and gamma-ray transition properties, which represent the evaluator's determination of the best values for these properties, based on all available experimental data. As of February 2008, the ENSDF database contains 16236 datasets for 3030 nuclides. (Taken from the NNDC's information page on ENSDF at http://www.nndc.bnl.gov/ensdf/ensdf_info.jsp) ENSDF may be browsed or the data may be retrieved based on nuclide, charge, or mass, or by indexed reaction and decay quantities. (Specialized interface)

  19. 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,...

  20. 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

    ... COMMISSION Configuration Management Plans for Digital Computer Software Used in Safety Systems of Nuclear... draft regulatory guide (DG), DG-1206, ``Configuration Management Plan for Digital Computer Software Used... Digital Computer Software Used in Safety Systems of Nuclear Power Plants'' is temporarily identified...

  1. 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,...

  2. 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,...

  3. 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,...

  4. 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,...

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. 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.

  14. 77 FR 20853 - Entergy Nuclear Operations, Inc.; Establishment of Atomic Safety and Licensing Board

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-04-06

    ... COMMISSION Entergy Nuclear Operations, Inc.; Establishment of Atomic Safety and Licensing Board Pursuant to..., notice is hereby given that an Atomic Safety and Licensing Board (Board) is being established to preside... the following administrative judges: Ann Marshall Young, Chair, Atomic Safety and Licensing...

  15. 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.

  16. 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.

  17. 76 FR 9351 - Patient Safety Organizations: Voluntary Delisting From West Virginia Center for Patient Safety

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-02-17

    ... Virginia State Medical. Association (WVSMA), of its status as a Patient Safety Organization (PSO). The Patient Safety and Quality Improvement Act of 2005 (Patient Safety Act), Public Law 109-41, 42 U.S.C. 299b... Patient Safety, a component entity of West Virginia Hospital Association, West Virginia Medical...

  18. Nuclear Industry Support Services by the Buffalo Materials Research Center

    SciTech Connect

    Henry, L.G. )

    1993-01-01

    The Buffalo Materials Research Center (BMRC) is located on the campus of the State University of New York at Buffalo, Principal facilities within BMRC include a 2-MW PULSTAR, low-enrichment reactor, an electron accelerator, and irradiated materials remote testing facilities. The reactor and the materials testing facilities have been utilized extensively in support of the power reactor community since 1961. This paper briefly highlights the nature and scope of this service. The BMRC is operated for the university by Buffalo Materials Research, Inc., a private for-profit company, which is a subsidiary of Materials Engineering Associates, Inc. (MEA), a Maryland-based materials testing company. A primary mission of MEA has been research on the effects of neutron irradiation on reactor structural materials, including those used for pressure vessel and piping systems. The combined resources of MEA and BMRC have played a pivotal role in the assessment of reactor pressure vessel safety both in the United States and abroad and in the development of new radiation-resistant steels.

  19. 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.

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

    SciTech Connect

    Saito, T.; Gasparini, M.

    2004-07-01

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

  1. 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.

  2. Cooperative Monitoring Center Occasional Paper/12: ENTNEA: A Concept for Enhancing Nuclear Transparency for Confidence Building in Northeast Asia

    SciTech Connect

    Nam, Man-Kwon; Shin, Sung-Tack

    1999-06-01

    Nuclear energy continues to be a strong and growing component of economic development in Northeast Asia. A broad range of nuclear energy systems already exists across the region and vigorous growth is projected. Associated with these capabilities and plans are various concerns about operational safety, environmental protection, and accumulation of spent fuel and other nuclear materials. We consider cooperative measures that might address these concerns. The confidence building measures suggested here center on the sharing of information to lessen concerns about nuclear activities or to solve technical problems. These activities are encompassed by an Enhanced Nuclear Transparency in Northeast Asia (ENTNEA) concept that would be composed of near-term, information-sharing activities and an eventual regional institution. The near-term activities would address specific concerns and build a tradition of cooperation; examples include radiation measurements for public safety and emergency response, demonstration of safe operations at facilities and in transportation, and material security in the back end of the fuel cycle. Linkages to existing efforts and organizations would be sought to maximize the benefits of cooperation. In the longer term, the new cooperative tradition might evolve into an ENTNEA institution. In institutional form, ENTNEA could combine the near-term activities and new cooperative activities, which might require an institutional basis, for the mutual benefit and security of regional parties.

  3. 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..

  4. 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

  5. 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.

  6. 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.

  7. Study of a conceptual nuclear energy center at Green River, Utah. Final summary report

    SciTech Connect

    Williams, J.T.

    1982-09-01

    This document summarizes a conceptual study on the feasibility and practicality of developing a nuclear energy center (NEC) at a representative Western site. The site selected for this conceptual study, an area of about 50 square miles, is located 15 miles south of Green River, Utah. The conceptual NEC would consist of nine nuclear electric generating units, arranged on the site in three clusters of three reactors each (triads), separated by about 2 1/2 miles. Of the total electric output of 11,250 MWe that the NEC could produce, about 82% is assumed to be transmitted out of Utah to Colorado, New Mexico, Arizona, Nevada, and California. The technical engineering issues studied included geology and seismology, plant design, low-level radioactive waste disposal, transmission, and construction schedules and costs. Socioeconomic issues included were demographics, land use, community service needs, and fiscal impacts. Environmental considerations included terrestrial and aquatic ecology, visual impact, and secondary population impacts. Radiological issues were concerned with the safety and risks of an NEC and an on-site low-level waste facility. Institutional issues included methods of ownership, taxation, implications of energy export, and water allocation. The basic finding was that an NEC would be technically feasible, but a number of socioeconomic and institutional issues would require resolution before a Western regional NEC could be considered a viable power plant siting option.

  8. 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.

  9. 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

  10. Advanced Outage and Control Center: Strategies for Nuclear Plant Outage Work Status Capabilities

    SciTech Connect

    Gregory Weatherby

    2012-05-01

    The research effort is a part of the Light Water Reactor Sustainability (LWRS) Program. LWRS is a research and development program sponsored by the Department of Energy, performed in close collaboration with industry to provide the technical foundations for licensing and managing the long-term, safe and economical operation of current nuclear power plants. The LWRS Program serves to help the US nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. The Outage Control Center (OCC) Pilot Project was directed at carrying out the applied research for development and pilot of technology designed to enhance safe outage and maintenance operations, improve human performance and reliability, increase overall operational efficiency, and improve plant status control. Plant outage management is a high priority concern for the nuclear industry from cost and safety perspectives. Unfortunately, many of the underlying technologies supporting outage control are the same as those used in the 1980’s. They depend heavily upon large teams of staff, multiple work and coordination locations, and manual administrative actions that require large amounts of paper. Previous work in human reliability analysis suggests that many repetitive tasks, including paper work tasks, may have a failure rate of 1.0E-3 or higher (Gertman, 1996). With between 10,000 and 45,000 subtasks being performed during an outage (Gomes, 1996), the opportunity for human error of some consequence is a realistic concern. Although a number of factors exist that can make these errors recoverable, reducing and effectively coordinating the sheer number of tasks to be performed, particularly those that are error prone, has the potential to enhance outage efficiency and safety. Additionally, outage management requires precise coordination of work groups that do not always share similar objectives. Outage

  11. 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.

  12. 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.

  13. General aviation crash safety program at Langley Research Center

    NASA Technical Reports Server (NTRS)

    Thomson, R. G.

    1976-01-01

    The purpose of the crash safety program is to support development of the technology to define and demonstrate new structural concepts for improved crash safety and occupant survivability in general aviation aircraft. The program involves three basic areas of research: full-scale crash simulation testing, nonlinear structural analyses necessary to predict failure modes and collapse mechanisms of the vehicle, and evaluation of energy absorption concepts for specific component design. Both analytical and experimental methods are being used to develop expertise in these areas. Analyses include both simplified procedures for estimating energy absorption capabilities and more complex computer programs for analysis of general airframe response. Full-scale tests of typical structures as well as tests on structural components are being used to verify the analyses and to demonstrate improved design concepts.

  14. 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.

  15. 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.

  16. 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.

  17. National Children's Center for Rural and Agricultural Health and Safety

    MedlinePlus

    ... Clinic Research Institute Careers Contact Us Patient Rights Privacy Location Our System Marshfield Clinic Marshfield Clinic Division of Education Marshfield Labs Security Health Plan Family Health Center Our Partners ... Website Privacy | Terms of Use | Non-discrimination Statement Copyright © 2012 - ...

  18. 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.

  19. 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.

  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. 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...

  2. Improving Performance of the System Safety Function at Marshall Space Flight Center

    NASA Technical Reports Server (NTRS)

    Kiessling, Ed; Tippett, Donald D.; Shivers, Herb

    2004-01-01

    The Columbia Accident Investigation Board (CAIB) determined that organizational and management issues were significant contributors to the loss of Space Shuttle Columbia. In addition, the CAIB observed similarities between the organizational and management climate that preceded the Challenger accident and the climate that preceded the Columbia accident. To prevent recurrence of adverse organizational and management climates, effective implementation of the system safety function is suggested. Attributes of an effective system safety program are presented. The Marshall Space Flight Center (MSFC) system safety program is analyzed using the attributes. Conclusions and recommendations for improving the MSFC system safety program are offered in this case study.

  3. 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.

  4. 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.

  5. 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...

  6. 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... Plants: LWR Edition.'' The new subsection is the Standard Review Plan (SRP), ``Introduction--Part 2... referenced. The SRP, subsection Introduction--Part 2 is under ADAMS Accession No. ML12142A237. NRC's PDR:...

  7. Federal Radiological Monitoring and Assessment Center Health and Safety Manual

    SciTech Connect

    FRMAC Health and Safety Working Group

    2012-03-20

    This manual is a tool to provide information to all responders and emergency planners and is suggested as a starting point for all organizations that provide personnel/assets for radiological emergency response. It defines the safety requirements for the protection of all emergency responders. The intent is to comply with appropriate regulations or provide an equal level of protection when the situation makes it necessary to deviate. In the event a situation arises which is not addressed in the manual, an appropriate management-level expert will define alternate requirements based on the specifics of the emergency situation. This manual is not intended to pertain to the general public.

  8. 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.

  9. 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.

  10. 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.

  11. 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)

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect

    M. D. Staiger M. C. Swenson

    2007-06-01

    This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

  18. 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.

  19. Nuclear Reaction Data File for Astrophysics (NRDF/A) in Hokkaido University Nuclear Reaction Data Center

    NASA Astrophysics Data System (ADS)

    Katō, Kiyoshi; Kimura, Masaaki; Furutachi, Naoya; Togashi, Tomoaki; Makinaga, Ayano; Otuka, Naohiko

    2010-06-01

    The activities of the Japan Nuclear Reaction Data Centre is explained. The main task of the centre is data compilation of Japanese nuclear reaction data in collaboration of the International Network of Nuclear Reaction Data Centres. As one of recent activities, preparation of a new database (NRDF/A) and evaluation of astronuclear reaction data are reported. Collaboration in the nuclear data activities among Asian countries is proposed.

  20. 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.

  1. 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

  2. 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.

  3. The Current Practices in Injury Prevention and Safety Helmet Use in an Air Force Medical Center

    DTIC Science & Technology

    2000-05-01

    Clinic at Malcom Grow Medical Center, Andrews Air Force Base, Maryland. Preventive counseling Preventing the occurrence of both mental and physical ...of their care. The primary care provider assumes ongoing responsibility for health maintenance and therapy for illness, including consultation with...PA) or a Medical Doctor (M.D.). Doctor of Osteopathy (D.O.), or Registered Nurse (R.N.). Safety helmet For the purpose of this study, the safety

  4. Collaborative development of a US/Russia safety information center database

    SciTech Connect

    Dvorack, M.A.; Smith, R.E.; Ananiychuk, V.N.; Volkova, N.A.

    1998-01-01

    One of the major outgrowths resulting from the collapse of the former Soviet Union (FSU) has been an increase in technical information exchange and dialogue between the Russian and American nuclear weapons laboratories. One area of such discussions is concerned with the safety of low probability, high consequence systems and operations. In order to further the understanding between the respective institutes in this important area, a collaborative effort has been established between Sandia National Laboratories and the two premier Russian nuclear weapons laboratories, Arzamas-16 and Chelyabinsk-70, in which a common database has been developed which contains safety information provided by all three laboratories. More than 1,200 documents have been placed by the three institutes into this database. This paper describes the details of this data base, including the types of safety information being stored.

  5. Embrittlement Database from the Radiation Safety Information Computational Center

    DOE Data Explorer

    The Embrittlement Data Base (EDB) is a comprehensive collection of data from surveillance capsules of U.S. commercial nuclear power reactors and from experiments in material test reactors. The collected data are contained in either the Power Reactor Embrittlement Data Base (PR-EDB) or the Test Reactor Embrittlement Data Base (TR-EDB). The EDB work includes verification of the quality of the EDB, provision for user-friendly software to access and process the data, exploration and/or confirmation of embrittlement prediction models, provision for rapid investigation of regulatory issues, and provision for the technical bases for voluntary consensus standards or regulatory guides. The EDB is designed for use with a personal computer. The data are collected into "raw data files." Traceability of all data is maintained by including complete references along with the page numbers. External data verification of the PR-EDB is the responsibility of the vendors, who were responsible for the insertion and testing of the materials in the surveillance capsules. Internal verification is accomplished by checking against references and checking for inconsistencies. Examples of information contained in the EDBs are: Charpy data, tensile data, reactor type, irradiation environments, fracture toughness data, instrumented Charpy data, pressure-temperature (P-T) data, chemistry data, and material history. The TR-EDB additionally has annealing Charpy data. The current version of the PR-EDB contains the test results from 269 Charpy capsules irradiated in 101 reactors. These results include 320 plate data points, 123 forging data points, 113 standard reference materials (SRMS) or correlation monitor (CM) points, 244 weld material data points, and 220 heat-affected-zone (HAZ) material data points. Similarly, the TR-EDB contains information for 290 SRM or CM points, 342 plate data points, 165 forging data points, 378 welds, and 55 HAZ materials. [copied from http://rsicc.ornl.gov/RelatedLinks.aspx?t=edb

  6. Probing an NV Center's Nuclear Spin Environment with Coherent Population Trapping

    NASA Astrophysics Data System (ADS)

    Levonian, David; Goldman, Michael; Singh, Swati; Markham, Matthew; Twitchen, Daniel; Lukin, Mikhail

    2016-05-01

    Nitrogen-vacancy (NV) centers in diamond have emerged as a versatile atom-like system, finding diverse applications in metrology and quantum information science, but interaction between the NV center's electronic spin and its nuclear spin environment represent a major source of decoherence. We use optical techniques to monitor and control the nuclear bath surrounding an NV center. Specifically, we create an optical Λ-system using the | +/- 1 > components of the NV center's spin-triplet ground state. When the Zeeman splitting between the two states is equal to the two-photon detuning between the lasers, population is trapped in the resulting dark state. Measuring the rate at which the NV center escapes from the dark state therefore gives information on how spin bath dynamics change the effective magnetic field experienced by the NV center. By monitoring statistics of the emitted photons, we plan to probe non-equilibrium dynamics of the bath.

  7. 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.

  8. 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)

  9. 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.

  10. 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.

  11. 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

  12. 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

  13. Safety, Reliability, and Quality Assurance Provisions for the Office of Aeronautics, Exploration and Technology Centers

    NASA Technical Reports Server (NTRS)

    1991-01-01

    This Handbook establishes general safety, reliability, and quality assurance (SR&QA) guidelines for use on flight and ground-based projects conducted at the Ames, Langley, and Lewis Research Centers, hereafter identified as the Office of Aeronautics, Exploration and Technology (OAET) Centers. This document is applicable to all projects and operations conducted at these Centers except for those projects covered by more restrictive provisions such as the Space Shuttle, Space Station, and unmanned spacecraft programs. This Handbook is divided into two parts. The first (Chapters 1 and 2) establishes the SR&QA guidelines applicable to the OAET Centers, and the second (Appendices A, B, C, and D) provides examples and definitions for the total SR&QA program. Each center should implement SR&QA programs using these guidelines with tailoring appropriate to the special projects conducted by each Center. This Handbook is issued in loose-leaf form and will be revised by page changes.

  14. 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).

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  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. 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.

  9. The roles and functions of a lunar base Nuclear Technology Center

    NASA Astrophysics Data System (ADS)

    Buden, D.; Angelo, J. A., Jr.

    This paper describes the roles and functions of a special Nuclear Technology Center which is developed as an integral part of a permanent lunar base. Numerous contemporary studies clearly point out that nuclear energy technology will play a major role in any successful lunar/Mars initiative program and in the overall establishment of humanity's solar system civilization. The key role of nuclear energy in the providing power has been recognized. A Nuclear Technology Center developed as part of a permanent lunar base can also help bring about many other nuclear technology applications, such as producing radioisotopes for self-illumination, food preservation, waste sterilization, and medical treatment; providing thermal energy for mining, materials processing and agricultural; and as a source of emergency habitat power. Designing such a center will involve the deployment, operation, servicing and waste product management and disposal of megawatt class reactor power plants. This challenge must be met with a minimum of direct human support at the facility. Furthermore, to support the timely, efficient integration of this Nuclear Technology Center in the evolving lunar base infrastructure, an analog of such a facility will be needed here on Earth.

  10. The roles and functions of a lunar base Nuclear Technology Center

    SciTech Connect

    Buden, D. ); Angelo, J.A. Jr. )

    1991-01-01

    This paper describes the roles and functions of a special Nuclear Technology Center which is developed as an integral part of a permanent lunar base. Numerous contemporary studies clearly point out that nuclear energy technology will play a major role in any successful lunar/Mars initiative program and in the overall establishment of humanity's solar system civilization. The key role of nuclear energy in the providing power has been recognized. A Nuclear Technology Center developed as part of of a permanent lunar base can also help bring about many other nuclear technology applications, such as producing radioisotopes for self-illumination, food preservation, waste sterilization, and medical treatment; providing thermal energy for mining, materials processing and agricultural; and as a source of emergency habitat power. Designing such a center will involve the deployment, operation, servicing and waste product management and disposal of megawatt class reactor power plants. This challenge must be met with a minimum of direct human support at the facility. Furthermore, to support the timely, efficient integration of this Nuclear Technology Center in the evolving lunar base infrastructure, an analog of such a facility will be needed here on Earth. 12 refs., 4 figs., 1 tab.

  11. 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.

  12. Cyclotron Production of Radionuclides for Nuclear Medicine at Academic Centers

    NASA Astrophysics Data System (ADS)

    Lapi, Suzanne

    2016-09-01

    The increase in use of radioisotopes for medical imaging has led to the development of new accelerator targetry and separation techniques for isotope production. For example, the development of longer-lived position emitting radionuclides has been explored to allow for nuclear imaging agents based on peptides, antibodies and nanoparticles. These isotopes (64Cu, 89Zr, 86Y) are typically produced via irradiation of solid targets on smaller cyclotrons (10-25 MeV) at academic or hospital based facilities. Recent research has further expanded the toolbox of PET tracers to include additional isotopes such as 52Mn, 55Co, 76Br and others. The smaller scale of these types of facilities can enable the straightforward involvement of students, thus adding to the next generation of nuclear science leaders. Research pertaining to development of robust and larger scale production technologies including solid target systems and remote systems for transport and purification of these isotopes has enabled both preclinical and clinical imaging research for many diseases. In particular, our group has focused on the use of radiolabeled antibodies for imaging of receptor expression in preclinical models and in a clinical trial of metastatic breast cancer patients.

  13. 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.

  14. The compliance of licensed US child care centers with national health and safety performance standards.

    PubMed Central

    Addiss, D G; Sacks, J J; Kresnow, M J; O'Neil, J; Ryan, G W

    1994-01-01

    The American Public Health Association and the American Academy of Pediatrics recently published health and safety guidelines for child care centers. A survey was conducted to determine the extent to which practices in US child care centers are reflective of these guidelines. Compliance with 16 guidelines ranged from 19.5% to 98.6%, varied considerably by state, and was not consistently associated with selected center characteristics. Prevention efforts should focus on practices for which compliance is low and on those that have the greatest disease- and injury-reducing potential. PMID:8017546

  15. Restructuring within an academic health center to support quality and safety: the development of the Center for Quality and Safety at the Massachusetts General Hospital.

    PubMed

    Bohmer, Richard M J; Bloom, Jonathan D; Mort, Elizabeth A; Demehin, Akinluwa A; Meyer, Gregg S

    2009-12-01

    Recent focus on the need to improve the quality and safety of health care has created new challenges for academic health centers (AHCs). Whereas previously quality was largely assumed, today it is increasingly quantifiable and requires organized systems for improvement. Traditional structures and cultures within AHCs, although well suited to the tripartite missions of teaching, research, and clinical care, are not easily adaptable to the tasks of measuring, reporting, and improving quality. Here, the authors use a case study of Massachusetts General Hospital's efforts to restructure quality and safety to illustrate the value of beginning with a focus on organizational culture, using a systematic process of engaging clinical leadership, developing an organizational framework dependent on proven business principles, leveraging focus events, and maintaining executive dedication to execution of the initiative. The case provides a generalizable example for AHCs of how applying explicit management design can foster robust organizational change with relatively modest incremental financial resources.

  16. 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.

  17. 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.

  18. 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.

  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. 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.

  1. 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.

  2. 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.

  3. 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.

  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. 77 FR 73541 - Safety Zone: Gilmerton Bridge Center Span Float-in, Elizabeth River; Norfolk, Portsmouth, and...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-11

    ... [Docket No. USCG-2012-0642] RIN 1625-AA00 Safety Zone: Gilmerton Bridge Center Span Float-in, Elizabeth... safety of life on navigable waters during the Gilmerton Bridge Center Span Float- in and bridge construction of span placement. This action is intended to restrict vessel traffic movement to protect...

  6. 77 FR 43557 - Safety Zone; Gilmerton Bridge Center Span Float-in, Elizabeth River; Norfolk, Portsmouth, and...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-25

    ... SECURITY Coast Guard 33 CFR Part 165 RIN 1625-AA00 Safety Zone; Gilmerton Bridge Center Span Float-in... provide for the safety of life on navigable waters during the Gilmerton Bridge Center Span Float- in and bridge construction of span placement. This action is intended to restrict vessel traffic movement...

  7. Nuclear magnetic resonance of external protons using continuous dynamical decoupling with shallow NV centers

    NASA Astrophysics Data System (ADS)

    de Las Casas, Charles; Ohno, Kenichi; Awschalom, David D.

    2015-03-01

    The nitrogen vacancy (NV) center in diamond is a paramagnetic defect with excellent spin properties that can reside within a few nanometers of the diamond surface, enabling atomic-scale magnetic resonance sensing of external nuclear spins. Here we use rotating frame longitudinal spin relaxation (T1ρ) based sensing schemes, known as Continuous Dynamical Decoupling (CDD), to detect external nuclear spins with shallow NV centers (<5 nm from the surface). Distinguishing neighboring nuclear spins from each other requires the NV center be near enough to create differences in the hyperfine shifts and coupling strengths of the nuclei. However, spin coherence time and consequently the sensitivity of dynamical decoupling techniques degrade sharply as NVs become shallower. We use strong continuous driving to overcome this fast decoherence and detect an ensemble of external nuclear spins using a single shallow NV center with a short T2 (<2 μs) at magnetic fields as high as 0.5 Tesla. The increased sensitivity of this method relative to pulsed dynamical decoupling techniques demonstrates the benefits of CDD for sensing with very shallow NV centers. This work was supported by DARPA, AFOSR, and the DIAMANT program.

  8. 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)

  9. 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.

  10. Nuclear Thermal Propulsion (NTP) Development Activities at the NASA Marshall Space Flight Center - 2006 Accomplishments

    NASA Technical Reports Server (NTRS)

    Ballard, Richard O.

    2007-01-01

    In 2005-06, the Prometheus program funded a number of tasks at the NASA-Marshall Space Flight Center (MSFC) to support development of a Nuclear Thermal Propulsion (NTP) system for future manned exploration missions. These tasks include the following: 1. NTP Design Develop Test & Evaluate (DDT&E) Planning 2. NTP Mission & Systems Analysis / Stage Concepts & Engine Requirements 3. NTP Engine System Trade Space Analysis and Studies 4. NTP Engine Ground Test Facility Assessment 5. Non-Nuclear Environmental Simulator (NTREES) 6. Non-Nuclear Materials Fabrication & Evaluation 7. Multi-Physics TCA Modeling. This presentation is a overview of these tasks and their accomplishments

  11. 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

    ... COMMISSION Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants..., ``Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants.'' This..., ``Maintenance and Inspection of Records.'' This RG is one of six RG revisions addressing computer...

  12. 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...

  13. The Occupational Safety of Health Professionals Working at Community and Family Health Centers

    PubMed Central

    Ozturk, Havva; Babacan, Elif

    2014-01-01

    Background: Healthcare professionals encounter many medical risks while providing healthcare services to individuals and the community. Thus, occupational safety studies are very important in health care organizations. They involve studies performed to establish legal, technical, and medical measures that must be taken to prevent employees from sustaining physical or mental damage because of work hazards. Objectives: This study was conducted to determine if the occupational safety of health personnel at community and family health centers (CHC and FHC) has been achieved. Martials and Methods: The population of this cross-sectional study comprised 507 nurses, 199 physicians, and 237 other medical personnel working at a total of 18 family health centers (FHC) and community health centers (CHC) in Trabzon, Turkey. The sample consisted of a total of 418 nurses, 156 physicians, and 123 other medical personnel. Sampling method was not used, and the researchers tried to reach the whole population. Data were gathered with the Occupational Safety Scale (OSS) and a questionnaire regarding demographic characteristics and occupational safety. Results: According to the evaluations of all the medical personnel, the mean ± SD of total score of the OSS was 3.57 ± 0.98; of the OSS’s subscales, the mean ± SD of the health screening and registry systems was 2.76 ± 1.44, of occupational diseases and problems was 3.04 ± 1.3 and critical fields control was 3.12 ± 1.62. In addition, occupational safety was found more insufficient by nurses (F = 14.18; P < 0.001). Conclusions: All healthcare personnel, particularly nurses working in CHCs and FHCs found occupational safety to be insufficient as related to protective and supportive activities. PMID:25558383

  14. 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.

  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.

  16. Decommissioning of the nuclear licensed facilities at the Fontenay aux Roses CEA center

    SciTech Connect

    Jeanjacques, Michel; Piketty, Laurence; Letuhaire, Nathalie; Mandard, Lionel; Meden, Igor; Estivie, David; Boissonneau, Jean Francois; Fouquereau, Alain; Pichereau, Eric; Binet, Cedric

    2007-07-01

    Available in abstract form only. Full text of publication follows: The French Atomic Energy Commission (CEA) center at Fontenay aux Roses (CEN-FAR) is the Commission's oldest center is located in the southern suburbs of Paris. It was opened on 26 March 1946 to host the first French nuclear reactor ZOE that went critical on 12 December 1946. The first laboratories were installed in existing buildings on the site. (authors)

  17. 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.

  18. 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.

  19. Nuclear Science References (NSR) from the National Nuclear Data Center (NNDC)

    DOE Data Explorer

    The Nuclear Science References (NSR) database is an indexed bibiliography of primary and secondary references in nuclear physics research. About 80 journals are regularly scanned for articles. Recent references are added on a weekly basis. Approximately 4300 entries are added to the database annually. In general, articles are included in NSR if they include measured, calculated, or deduced quantitative nuclear structure or reaction data. Papers that apply previously known data are generally not included. Examples of this include neutron activation analysis using known cross sections or radiological dating using known half-lives. The database can be searched like a normal bibliographic database but can also be searched by the data that distinguishes it, data such as the nuclide, target/parent/daughter, reaction, incident particles, and outgoing particles. (Specialized Interface) [Taken from the NSR Help pages at http://www.nndc.bnl.gov/nsr/nsr_help.jsp

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. Nuclear energy related research

    NASA Astrophysics Data System (ADS)

    Rintamaa, R.

    1992-05-01

    The annual Research Program Plan describes publicly funded nuclear energy related research to be carried out mainly at the Technical Research Center of Finland (VTT) in 1992. The research is financed primarily by the Ministry of Trade and Industry (KTM), the Finnish Center for Radiation and Nuclear Safety (STUK), and VTT itself. Other research institutes, utilities, and industry also contribute to many projects.

  6. Radiotherapy and Nuclear Medicine Project for an Integral Oncology Center at the Oaxaca High Specialization Regional Hospital

    NASA Astrophysics Data System (ADS)

    De Jesús, M.; Trujillo-Zamudio, F. E.

    2010-12-01

    A building project of Radiotherapy & Nuclear Medicine services (diagnostic and therapy), within an Integral Oncology Center (IOC), requires interdisciplinary participation of architects, biomedical engineers, radiation oncologists and medical physicists. This report focus on the medical physicist role in designing, building and commissioning stages, for the final clinical use of an IOC at the Oaxaca High Specialization Regional Hospital (HRAEO). As a first step, during design stage, the medical physicist participates in discussions about radiation safety and regulatory requirements for the National Regulatory Agency (called CNSNS in Mexico). Medical physicists propose solutions to clinical needs and take decisions about installing medical equipment, in order to fulfill technical and medical requirements. As a second step, during the construction stage, medical physicists keep an eye on building materials and structural specifications. Meanwhile, regulatory documentation must be sent to CNSNS. This documentation compiles information about medical equipment, radioactivity facility, radiation workers and nuclear material data, in order to obtain the license for the linear accelerator, brachytherapy and nuclear medicine facilities. As a final step, after equipment installation, the commissioning stage takes place. As the conclusion, we show that medical physicists are essentials in order to fulfill with Mexican regulatory requirements in medical facilities.

  7. Radiotherapy and Nuclear Medicine Project for an Integral Oncology Center at the Oaxaca High Specialization Regional Hospital

    SciTech Connect

    De Jesus, M.; Trujillo-Zamudio, F. E.

    2010-12-07

    A building project of Radiotherapy and Nuclear Medicine services (diagnostic and therapy), within an Integral Oncology Center (IOC), requires interdisciplinary participation of architects, biomedical engineers, radiation oncologists and medical physicists. This report focus on the medical physicist role in designing, building and commissioning stages, for the final clinical use of an IOC at the Oaxaca High Specialization Regional Hospital (HRAEO). As a first step, during design stage, the medical physicist participates in discussions about radiation safety and regulatory requirements for the National Regulatory Agency (called CNSNS in Mexico). Medical physicists propose solutions to clinical needs and take decisions about installing medical equipment, in order to fulfill technical and medical requirements. As a second step, during the construction stage, medical physicists keep an eye on building materials and structural specifications. Meanwhile, regulatory documentation must be sent to CNSNS. This documentation compiles information about medical equipment, radioactivity facility, radiation workers and nuclear material data, in order to obtain the license for the linear accelerator, brachytherapy and nuclear medicine facilities. As a final step, after equipment installation, the commissioning stage takes place. As the conclusion, we show that medical physicists are essentials in order to fulfill with Mexican regulatory requirements in medical facilities.

  8. 78 FR 4404 - DOE Response to Recommendation 2012-2 of the Defense Nuclear Facilities Safety Board, Hanford...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-22

    ... Flammable Gas Safety Strategy AGENCY: Department of Energy. ACTION: Notice. SUMMARY: On September 28, 2012... Farms Flammable Gas Safety Strategy, to the Department of Energy. In accordance with section 315(b) of... Nuclear Facilities Safety Board (Board) Recommendation 2012-2, Hanford Tank Farms Flammable Gas...

  9. 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,...

  10. Noise-Resilient Quantum Computing with a Nitrogen-Vacancy Center and Nuclear Spins.

    PubMed

    Casanova, J; Wang, Z-Y; Plenio, M B

    2016-09-23

    Selective control of qubits in a quantum register for the purposes of quantum information processing represents a critical challenge for dense spin ensembles in solid-state systems. Here we present a protocol that achieves a complete set of selective electron-nuclear gates and single nuclear rotations in such an ensemble in diamond facilitated by a nearby nitrogen-vacancy (NV) center. The protocol suppresses internuclear interactions as well as unwanted coupling between the NV center and other spins of the ensemble to achieve quantum gate fidelities well exceeding 99%. Notably, our method can be applied to weakly coupled, distant spins representing a scalable procedure that exploits the exceptional properties of nuclear spins in diamond as robust quantum memories.

  11. Development of Risk Assessment Matrix for NASA Engineering and Safety Center

    NASA Technical Reports Server (NTRS)

    Malone, Roy W., Jr.; Moses, Kelly

    2004-01-01

    This paper describes a study, which had as its principal goal the development of a sufficiently detailed 5 x 5 Risk Matrix Scorecard. The purpose of this scorecard is to outline the criteria by which technical issues can be qualitatively and initially prioritized. The tool using this score card has been proposed to be one of the information resources the NASA Engineering and Safety Center (NESC) takes into consideration when making decisions with respect to incoming information on safety concerns across the entire NASA agency. The contents of this paper discuss in detail each element of the risk matrix scorecard, definitions for those elements and the rationale behind the development of those definitions. This scorecard development was performed in parallel with the tailoring of the existing Futron Corporation Integrated Risk Management Application (IRMA) software tool. IRMA was tailored to fit NESC needs for evaluating incoming safety concerns and was renamed NESC Assessment Risk Management Application (NAFMA) which is still in developmental phase.

  12. NASA Engineering and Safety Center (NESC) Enhanced Melamine (ML) Foam Acoustic Test (NEMFAT)

    NASA Technical Reports Server (NTRS)

    McNelis, Anne M.; Hughes, William O.; McNelis, Mark E.

    2014-01-01

    The NASA Engineering and Safety Center (NESC) funded a proposal to achieve initial basic acoustic characterization of ML (melamine) foam, which could serve as a starting point for a future, more comprehensive acoustic test program for ML foam. A project plan was developed and implemented to obtain acoustic test data for both normal and enhanced ML foam. This project became known as the NESC Enhanced Melamine Foam Acoustic Test (NEMFAT). This document contains the outcome of the NEMFAT project.

  13. 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.

  14. 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.

  15. Experimental Unevaluated Nuclear Data List (XUNDL) from the National Nuclear Data Center (NNDC)

    DOE Data Explorer

    XUNDL was established in 1998. While it contains compiled nuclear structure data in the "ENSDF" format, the difference is that the ENSDF collection holds evaluated data, while XUNDL holds experimental data that have not yet been evaluated. The manual for one of the evaluated collections defines evaluation as "the process of analyzing experimentally measured cross-section data, combining them with the predictions of nuclear model calculations, and attempting to extract the true value of a cross section." (Document ENDF-102) The XUNDL data are the experimental data before these calculations and comparisons are done. In general, the information in a given XUNDL dataset comes from a single journal article or from a set of closely related articles by one group of authors. The bulk of the compilations are carried out by the Nuclear Data Group at McMaster University. As of February 2008, the XUNDL database contained more than 2500 datasets for 1484 nuclides. The data can be searched through specialized interfaces or browsed. There is also a special list of hypernucleus datasets available. (Specialized Interface)

  16. 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.

  17. 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.

  18. Pilot program to identify valve failures which impact the safety and operation of light water nuclear power plants

    SciTech Connect

    Tsacoyeanes, J. C.; Raju, P. P.

    1980-04-01

    The pilot program described has been initiated under the Department of Energy Light Water Reactor Safety Research and Development Program and has the following specific objectives: to identify the principal types and causes of failures in valves, valve operators and their controls and associated hardware, which lead to, or could lead to plant trip; and to suggest possible remedies for the prevention of these failures and recommend future research and development programs which could lead to minimizing these valve failures or mitigating their effect on plant operation. The data surveyed cover incidents reported over the six-year period, beginning 1973 through the end of 1978. Three sources of information on valve failures have been consulted: failure data centers, participating organizations in the nuclear power industry, and technical documents.

  19. 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

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. Information Strategy of Nuclear Training Center Ljubljana in the Area of Radioactive Waste Management

    SciTech Connect

    Jeneie, I.

    2008-07-01

    Slovenia has plans to build a repository for low- and medium-radioactive waste by 2013, the location in the very neighborhood of nuclear power plant is almost chosen, but the final approval hasn't been granted yet. The main obstacle is public opinion. Public information activities are therefore vitally important. One of the most important players in this area in Slovenia is Nuclear Training Center in Ljubljana. Though its main task is training of nuclear professionals, it has a significant role in dissemination of knowledge about radioactivity and nuclear technology also among general public. Public information is focused on youngsters. Almost one half of every generation of schoolchildren in Slovenia visits the Information center yearly and in May 2007, we have celebrated the 100,000. visitor since its opening. Live lectures, exhibition, publications and laboratory demonstrations are offered. To measure the opinion of youngsters about nuclear power and get a feed-back for our activities about 1000 youngsters are polled every year since 1993 using the same basic set of questions. The paper describes the information strategy, types of lectures and information materials, permanent exhibition with the most important exhibits. Furthermore, the results of yearly polls of our visitors and comparison with relevant Euro-barometer polls are presented. (authors)

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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

  16. 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.

  17. Energy Frontier Research Center Center for Materials Science of Nuclear Fuels

    SciTech Connect

    Todd Allen

    2014-04-01

    Scientific Successes • The first phonon density of states (PDOS) measurements for UO2 to include anharmonicity were obtained using time-of-flight inelastic neutron scattering at the Spallation Neutron Source (SNS), and an innovative, experimental-based anharmonic smoothing technique has enabled quantitative benchmarking of ab initio PDOS simulations. • Direct comparison between anharmonicity-smoothed ab initio PDOS simulations for UO2 and experimental measurements has demonstrated the need for improved understanding of UO2 at the level of phonon dispersion, and, further, that advanced lattice dynamics simulations including finite temperatures approaches will be required for handling this strongly correlated nuclear fuel. • PDOS measurements performed on polycrystalline samples have identified the phonon branches and energy ranges most highly impacted by fission-product and hyper-stoichiometry lattice defects in UO2. These measurements have revealed the broad-spectrum impact of oxygen hyper-stoichiometry on thermal transport. The reduction in thermal conductivity caused by hyper-stoichiometry is many times stronger than that caused by substitutional fission-product impurities. • Laser-based thermo-reflectance measurements on UO2 samples irradiated with light (i.e. He) ions to introduce point defects have been coupled with MD simulations and lattice parameter measurements to determine the role of uranium and oxygen point defects in reducing thermal conductivity. • A rigorous perturbation theory treatment of phonon lifetimes in UO2 based on a 3D discretization of the Brillouin zone coupled with experimentally measured phonon dispersion has been implemented that produces improved predictions of the temperature dependent thermal conductivity. • Atom probe investigations of the influence of grain boundary structure on the segregation behavior of Kr in UO2 have shown that smaller amounts of Kr are present at low angle grain boundaries than at large angle grain

  18. Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect

    Staiger, Merle Daniel; M. C. Swenson

    2005-01-01

    This report documents an inventory of calcined waste produced at the Idaho Nuclear Technology and Engineering Center during the period from December 1963 to May 2000. The report was prepared based on calciner runs, operation of the calcined solids storage facilities, and miscellaneous operational information that establishes the range of chemical compositions of calcined waste stored at Idaho Nuclear Technology and Engineering Center. The report will be used to support obtaining permits for the calcined solids storage facilities, possible treatment of the calcined waste at the Idaho National Engineering and Environmental Laboratory, and to ship the waste to an off-site facility including a geologic repository. The information in this report was compiled from calciner operating data, waste solution analyses and volumes calcined, calciner operating schedules, calcine temperature monitoring records, and facility design of the calcined solids storage facilities. A compact disk copy of this report is provided to facilitate future data manipulations and analysis.

  19. Socioeconomic impacts: study of a conceptual nuclear energy center at Green River, Utah

    SciTech Connect

    Weaver, R.; Taylor, J.; Burnett, K.; Greenberg, B.

    1982-02-01

    This document constitutes a segment of a feasibility study investigating the ramifications of constructing a nuclear energy center (NEC) in an arid western region. In this phase of the study, the impacts on socioeconomic conditions in the surrounding communities and possible ways of financing and mitigating these impacts were examined. The general conclusion reached is that the socioeconomic impacts of a nuclear energy center in the Green River area of Southeastern Utah would not impose an absolute bar to NEC development. The economy of the NEC impact area would be substantially transformed by the NEC. In particular, Green River city itself would change from its current status as a relatively stable rural economy with an agricultural, mining, and recreation base to a major city with over 20,000 permanent relatively high income residents. The NEC, by itself, would provide a tax base more than adequate to finance required expansion of public facilities and public human service provisions.

  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. Anomaly Analysis: NASA's Engineering and Safety Center Checks Recurring Shuttle Glitches

    NASA Technical Reports Server (NTRS)

    Morring, Frank, Jr.

    2004-01-01

    The NASA Engineering and Safety Center (NESC), set up in the wake of the Columbia accident to backstop engineers in the space shuttle program, is reviewing hundreds of recurring anomalies that the program had determined don't affect flight safety to see if in fact they might. The NESC is expanding its support to other programs across the agency, as well. The effort, which will later extend to the International Space Station (ISS), is a principal part of the attempt to overcome the normalization of deviance--a situation in which organizations proceeded as if nothing was wrong in the face of evidence that something was wrong--cited by sociologist Diane Vaughn as contributing to both space shuttle disasters.

  2. Graphical user interfaces for McCellan Nuclear Radiation Center (MNRC).

    SciTech Connect

    Brown-VanHoozer, S. A.

    1998-08-27

    McClellan's Nuclear Radiation Center (MNRC) control console is in the process of being replaced due to spurious scrams, outdated software, and obsolete parts. The intent of the new control console is to eliminate the existing problems by installing a UNIX-based computer system with industry-standard interface software and incorporating human factors during all stages of the graphical user interface (GUI) development and control console design.

  3. University Reactor Conversion Lessons Learned Workshop for Texas A&M University Nuclear Science Center Reactor

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Meyer

    2007-04-01

    The objectives of this meeting were to capture the observations, insights, issues, concerns, and ideas of those involved in the Texas A&M University Nuclear Science Center (TAMU NSC) TRIGA Reactor Conversion so that future efforts can be conducted with greater effectiveness, efficiency, and with fewer challenges. This workshop was held in conjunction with a similar workshop for the University of Florida Reactor Conversion. Some of the generic lessons from that workshop are included in this report for completeness.

  4. Strongly polarizing weakly coupled 13C nuclear spins with optically pumped nitrogen-vacancy center

    PubMed Central

    Wang, Ping; Liu, Bao; Yang, Wen

    2015-01-01

    Enhancing the polarization of nuclear spins surrounding the nitrogen-vacancy (NV) center in diamond has recently attracted widespread attention due to its various applications. Here we present an analytical formula that not only provides a clear physical picture for the recently observed polarization reversal of strongly coupled13C nuclei over a narrow range of magnetic field [H. J. Wang et al., Nat. Commun. 4, 1940 (2013)], but also demonstrates the possibility to strongly polarize weakly coupled13C nuclei. This allows sensitive magnetic field control of the 13C nuclear spin polarization for NMR applications and significant suppression of the 13C nuclear spin noise to prolong the NV spin coherence time. PMID:26521962

  5. Evidence of the radioactive fallout in the center of Asia (Russia) following the Fukushima Nuclear Accident.

    PubMed

    Bolsunovsky, A; Dementyev, D

    2011-11-01

    It was recently reported that radioactive fallout due to the Fukushima Nuclear Accident was detected in environmental samples collected in the USA and Greece, which are very far away from Japan. In April-May 2011, fallout radionuclides ((134)Cs, (137)Cs, (131)I) released in the Fukushima Nuclear Accident were detected in environmental samples at the city of Krasnoyarsk (Russia), situated in the center of Asia. Similar maximum levels of (131)I and (137)Cs/(134)Cs and (131)I/(137)Cs ratios in water samples collected in Russia and Greece suggest the high-velocity movement of the radioactive contamination from the Fukushima Nuclear Accident and the global effects of this accident, similar to those caused by the Chernobyl accident.

  6. 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).

  7. An Evaluation of North Korea’s Nuclear Test by Belbasi Nuclear Tests Monitoring Center-KOERI

    NASA Astrophysics Data System (ADS)

    Necmioglu, O.; Meral Ozel, N.; Semin, K.

    2009-12-01

    Bogazici University and Kandilli Observatory and Earthquake Research Institute (KOERI) is acting as the Turkish National Data Center (NDC) and responsible for the operation of the International Monitoring System (IMS) Primary Seismic Station (PS-43) under Belbasi Nuclear Tests Monitoring Center for the verification of compliance with the Comprehensive Nuclear-Test-Ban Treaty (CTBT) since February 2000. The NDC is responsible for operating two arrays which are part of the IMS, as well as for transmitting data from these stations to the International Data Centre (IDC) in Vienna. The Belbasi array was established in 1951, as a four-element (Benioff 1051) seismic array as part of the United States Atomic Energy Detection System (USAEDS). Turkish General Staff (TGS) and U.S. Air Force Technical Application Center (AFTAC) under the Defense and Economic Cooperation Agreement (DECA) jointly operated this short period array. The station was upgraded and several seismometers were added to array during 1951 and 1994 and the station code was changed from BSRS (Belbasi Seismic Research Station) to BRTR-PS43 later on. PS-43 is composed of two sub-arrays (Ankara and Keskin): the medium-period array with a ~40 km radius located in Ankara and the short-period array with a ~3 km radius located in Keskin. Each array has a broadband element located at the middle of the circular geometry. Short period instruments are installed at depth 30 meters from the surface while medium and broadband instruments are installed at depth 60 meters from surface. On 25 May 2009, The Democratic People’s Republic of Korea (DPRK) claimed that it had conducted a nuclear test. Corresponding seismic event was recorded by IMS and IDC released first automatic estimation of time (00:54:43 GMT), location (41.2896°N and 129.0480°E) and the magnitude (4.52 mb) of the event in less than two hours time (USGS: 00:54:43 GMT; 41.306°N, 129.029°E; 4.7 mb) During our preliminary analysis of the 25th May 2009 DPRK

  8. How the Space Data Center Is Improving Safety of Space Operations

    NASA Astrophysics Data System (ADS)

    Kelso, T. S.

    2010-09-01

    In an effort to mitigate the risks associated with satellite close approaches in the geostationary belt, satellite operators began to come together in early 2008 to establish a prototype GEO data center. That prototype provided a framework for operators to share orbital data for their fleets to be used to perform conjunction analysis and provide automated notifications of close approaches via the SOCRATES-GEO service. That service was extended to LEO operations in mid-2009 and, as of early 2010, the prototype was supporting 20 operators from over a dozen countries by automatically screening 300 satellites for close approaches twice each day. In April 2010, the prototype data center operated by the Center for Space Standards & Innovation (CSSI) was a key reason AGI was selected by the Space Data Association (SDA) to develop the SDA’s new Space Data Center (SDC). This paper will address how the SDC will use a service-oriented architecture (SOA) to support orbital operations by increasing the efficiency of analysis to mitigate the risk of conjunctions and radio frequency interference, thereby enhancing overall safety of flight.

  9. Purpose, Principles, and Challenges of the NASA Engineering and Safety Center

    NASA Technical Reports Server (NTRS)

    Gilbert, Michael G.

    2016-01-01

    NASA formed the NASA Engineering and Safety Center in 2003 following the Space Shuttle Columbia accident. It is an Agency level, program-independent engineering resource supporting NASA's missions, programs, and projects. It functions to identify, resolve, and communicate engineering issues, risks, and, particularly, alternative technical opinions, to NASA senior management. The goal is to help ensure fully informed, risk-based programmatic and operational decision-making processes. To date, the NASA Engineering and Safety Center (NESC) has conducted or is actively working over 600 technical studies and projects, spread across all NASA Mission Directorates, and for various other U.S. Government and non-governmental agencies and organizations. Since inception, NESC human spaceflight related activities, in particular, have transitioned from Shuttle Return-to-Flight and completion of the International Space Station (ISS) to ISS operations and Orion Multi-purpose Crew Vehicle (MPCV), Space Launch System (SLS), and Commercial Crew Program (CCP) vehicle design, integration, test, and certification. This transition has changed the character of NESC studies. For these development programs, the NESC must operate in a broader, system-level design and certification context as compared to the reactive, time-critical, hardware specific nature of flight operations support.

  10. 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... regulatory guide (DG), DG-1208, ``Software Unit Testing for Digital Computer Software used in Safety Systems... revision endorses, with clarifications, the enhanced consensus practices for testing of computer...

  11. 77 FR 50720 - Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-08-22

    ... COMMISSION Test Documentation for Digital Computer Software Used in Safety Systems of Nuclear Power Plants... regulatory guide (DG), DG-1207, ``Test Documentation for Digital Computer Software used in Safety Systems of... software and computer systems as described in the Institute of Electrical and Electronics Engineers...

  12. 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....

  13. 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....

  14. 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.

  15. 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.

  16. 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.

  17. 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)

  18. 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.

  19. Safety and Environment- Masterplan 2020 of DLR's Rocket Test Center Lampoldhausen

    NASA Astrophysics Data System (ADS)

    Haberzettl, Andreas; Dommers, Michael

    2013-09-01

    The German Aerospace Center DLR is the German research institute with approximately 7000 employees in 16 domestic locations. Among the research priorities of the German Aerospace Center DLR includes aerospace, energy and transport. DLR is institutionally supported by federal and state governments.Next funding sources arise in the context of third-party funds business (contract research and public contracts and subsidiaries). Main activities of the test center Lampoldshausen are testing of ARIANE's main and upper stage engines in the frame of ESA contracts.In the last years the test center of the DLR in Lampoldshausen has grown strongly, so that the number of employees is actually of about 230. The testing department is mainly responsible for rocket combustion testing according to customer requirements.Two kinds of test facilities are operated, sea level test benches and the altitude simulation test facilities.In addition to the DLR's growth also the activities of the industrial partner ASTRIUM has been elevated so that actually nearly 600 employees are present on site Lampoldshausen.The management of the site in relation to safety and security requires special measures with special respect to the presence of more people inside the testing area in order to guarantee trouble-free and safe experimental operation onsite the DLR's test plants. In order to meet with the future needs of continuing growth, the security and safety requirements have to be adopted.This report gives comprehensive outlook information about future possible scenarios of our coming tasks.Main driving force for future requests is the evolution of the rocket ARIANE. The testing of the new upper stage test facility for ARIANE 5 midlife evolution has been started. A new test position P5.2 is foreseen to perform the qualification of the new upper stage with the VINCI engine. This project will be very complex, in parallel running operation processes will require special procedures related to the overall

  20. 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

  1. 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.

  2. 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.

  3. 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.''

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Materials characterization center workshop on compositional and microstructural analysis of nuclear waste materials. Summary report

    SciTech Connect

    Daniel, J.L.; Strachan, D.M.; Shade, J.W.; Thomas, M.T.

    1981-06-01

    The purpose of the Workshop on Compositional and Microstructural Analysis of Nuclear Waste Materials, conducted November 11 and 12, 1980, was to critically examine and evaluate the various methods currently used to study non-radioactive, simulated, nuclear waste-form performance. Workshop participants recognized that most of the Materials Characterization Center (MCC) test data for inclusion in the Nuclear Waste Materials Handbook will result from application of appropriate analytical procedures to waste-package materials or to the products of performance tests. Therefore, the analytical methods must be reliable and of known accuracy and precision, and results must be directly comparable with those from other laboratories and from other nuclear waste materials. The 41 participants representing 18 laboratories in the United States and Canada were organized into three working groups: Analysis of Liquids and Solutions, Quantitative Analysis of Solids, and Phase and Microstructure Analysis. Each group identified the analytical methods favored by their respective laboratories, discussed areas needing attention, listed standards and reference materials currently used, and recommended means of verifying interlaboratory comparability of data. The major conclusions from this workshop are presented.

  12. 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.

  13. A survey of nuclear-related agreements and possibilities for nuclear cooperation in South Asia: Cooperative Monitoring Center Occasional Paper/15

    SciTech Connect

    RAJEN,GAURAV

    2000-04-01

    Several existing nuclear-related agreements already require India and Pakistan, as members, to share information. The agreements are bilateral, regional, and international. Greater nuclear transparency between India and Pakistan could be promoted by first understanding the information flows required by existing agreements. This understanding is an essential step for developing projects that can incrementally advance the sensitivity of the information being shared. This paper provides a survey of existing nuclear-related agreements involving India and Pakistan, and suggests future confidence-building projects using the frameworks provided by these agreements. The Bilateral Agreement on the Prohibition of Attack against Nuclear Reactors and Nuclear Facilities is discussed as a basis for creating further agreements on restricting the use and deployment of nuclear weapons. The author suggests options for enhancing the value of the list of nuclear facilities exchanged annually as a part of this agreement. The International Atomic Energy Agency's regional cooperation agreement among countries in the Asia-Pacific region is an opportunity for greater subregional nuclear cooperation in South Asia. Linking the regional agreement with South Asian environmental cooperation and marine pollution protection efforts could provide a framework for projects involving Indian and Pakistani coastal nuclear facilities. Programs of the Food and Agriculture Organization of the United Nations that use nuclear techniques to increase food and crop production and optimize water management in arid areas also provide similar opportunities for nuclear cooperation. Other frameworks for nuclear cooperation originate from international conventions related to nuclear safety, transportation of nuclear wastes, worker protection against ionizing radiation, and the nondeployment of nuclear weapons in certain areas. The information shared by existing frameworks includes: laws and regulations (including

  14. HI Clouds Near the Galactic Center: Possible Tracers of the Nuclear Wind

    NASA Astrophysics Data System (ADS)

    Lockman, Felix J.; McClure-Griffiths, Naomi; DiTeodoro, Enrico

    2017-01-01

    We have used the Green Bank Telescope to discover more than one hundred neutral hydrogen clouds that appear to be embedded in the Fermi Bubble -- the Milky Way’s nuclear wind. With the other members of this population that were previously found with the Australia Telescope Compact Array, we now have a sample of about 200 such clouds. They are identified by their peculiar velocities. The cloud kinematics show no trace of Galactic rotation or association with the Galactic bar. Near longitude zero the clouds can have values of VLSR = +-200 km/s. No clouds have been detected with |VLSR| > 350 km/s. The clouds are concentrated toward the Galactic plane, but some are still found to |b|=10 degrees, or z > 1 kpc at the Galactic Center, where the current surveys end. These clouds are important tracers of conditions in the nuclear wind of the Milky Way.

  15. Final report. Conceptual studies nuclear energy center Lake Hartwell, S. C. , Phase III

    SciTech Connect

    Not Available

    1981-01-01

    This document summarizes a conceptual study on the feasibility and practicality of developing a nuclear energy center (NEC) at a specific site in the SSEB region. The site selected for this conceptual study is at Lake Hartwell, South Carolina. The conceptual NEC at Lake Hartwell consists of twelve nuclear electric generating units, arranged on the site in four clusters of three units each, known as triads. The nominal distance between triads was selected as 2-1/2 miles. Each unit was assumed to be a 1250 MW(e). The total electric output of 15,000 MWe would be transmitted to five major utilities in South Carolina, North Carolina, and Georgia. The basic finding was that the concept of a NEC on the Lake Hartwell site is feasible, but further analysis of institutional issues and possible legislation would be required.

  16. Historical perspectives - The role of the NASA Lewis Research Center in the national space nuclear power programs

    NASA Technical Reports Server (NTRS)

    Bloomfield, H. S.; Sovie, R. J.

    1991-01-01

    The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many natural space nuclear power and propulsion programs.

  17. Historical perspectives: The role of the NASA Lewis Research Center in the national space nuclear power programs

    NASA Technical Reports Server (NTRS)

    Bloomfield, H. S.; Sovie, R. J.

    1991-01-01

    The history of the NASA Lewis Research Center's role in space nuclear power programs is reviewed. Lewis has provided leadership in research, development, and the advancement of space power and propulsion systems. Lewis' pioneering efforts in nuclear reactor technology, shielding, high temperature materials, fluid dynamics, heat transfer, mechanical and direct energy conversion, high-energy propellants, electric propulsion and high performance rocket fuels and nozzles have led to significant technical and management roles in many national space nuclear power and propulsion programs.

  18. 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.

  19. 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

  20. Electron-Mediated Nuclear-Spin Interactions between Distant Nitrogen-Vacancy Centers

    NASA Astrophysics Data System (ADS)

    Bermudez, A.; Jelezko, F.; Plenio, M. B.; Retzker, A.

    2011-10-01

    We propose a scheme enabling controlled quantum coherent interactions between separated nitrogen-vacancy centers in diamond in the presence of strong magnetic fluctuations. The proposed scheme couples nuclear qubits employing the magnetic dipole-dipole interaction between the electron spins and, crucially, benefits from the suppression of the effect of environmental magnetic field fluctuations thanks to a strong microwave driving. This scheme provides a basic building block for a full-scale quantum-information processor or quantum simulator based on solid-state technology.

  1. Idaho Nuclear Technology and Engineering Center Newly Generated Liquid Waste Demonstration Project Feasibility Study

    SciTech Connect

    Herbst, A.K.

    2000-02-01

    A research, development, and demonstration project for the grouting of newly generated liquid waste (NGLW) at the Idaho Nuclear Technology and Engineering Center is considered feasible. NGLW is expected from process equipment waste, decontamination waste, analytical laboratory waste, fuel storage basin waste water, and high-level liquid waste evaporator condensate. The potential grouted waste would be classed as mixed low-level waste, stabilized and immobilized to meet RCRA LDR disposal in a grouting process in the CPP-604 facility, and then transported to the state.

  2. 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.

  3. Two-center interferences and nuclear wave packet imaging in dissociating H2+ molecule

    NASA Astrophysics Data System (ADS)

    Picon, Antonio; Bahabad, Alon; Kapteyn, Henry C.; Murnane, Margaret M.; Becker, Andreas

    2011-05-01

    Double-slit like interferences similar to those observed by Young in his experiment with light appear also in the photoionization of diatomic molecules. The partial electron waves ejected from the two atomic centers of the molecule take the role of the coherent light waves emerging from the two holes in Youngs experiment. We analyze theoretically and numerically a pump-probe scenario with two attosecond pulses in the hydrogen molecular ion. The first attosecond pulse induces the dissociation of the molecule, the second attosecond pulse is ionizing the molecule. By varying the delay between the pump and probe pulses we show how the two-center interferences allow to image main features of the nuclear wave packet, namely its velocity, internuclear distance, and spreading. Supported by Postdoctoral Program of the Spanish Government and NSF.

  4. Radiation Safety System of the B-Factory at the Stanford Linear Accelerator Center

    SciTech Connect

    Liu, James C.

    1998-10-12

    The radiation safety system (RSS) of the B-Factory accelerator facility at the Stanford Linear Accelerator Center (SLAC) is described. The RSS, which is designed to protect people from prompt radiation exposure due to beam operation, consists of the access control system (ACS) and the radiation containment system (RCS). The ACS prevents people from being exposed to the very high radiation levels inside a beamline shielding housing. The ACS consists of barriers, a standard entry module at every entrance, and beam stoppers. The RCS prevents people from being exposed to the radiation outside a shielding housing, due to either normal or abnormal operation. The RCS consists of power limiting devices, shielding, dump/collimator, and an active radiation monitor system. The inter-related system elements for the ACS and RCS, as well as the associated interlock network, are described. The policies and practices in setting up the RSS are also compared with the regulatory requirements.

  5. Environment, safety and health compliance assessment, Feed Materials Production Center, Fernald, Ohio

    SciTech Connect

    Not Available

    1989-09-01

    The Secretary of Energy established independent Tiger Teams to conduct environment, safety, and health (ES H) compliance assessments at US Department of Energy (DOE) facilities. This report presents the assessment of the Feed Materials Production Center (FMPC) at Fernald, Ohio. The purpose of the assessment at FMPC is to provide the Secretary with information regarding current ES H compliance status, specific ES H noncompliance items, evaluation of the adequacy of the ES H organizations and resources (DOE and contractor), and root causes for noncompliance items. Areas reviewed included performance under Federal, state, and local agreements and permits; compliance with Federal, state and DOE orders and requirements; adequacy of operations and other site activities, such as training, procedures, document control, quality assurance, and emergency preparedness; and management and staff, including resources, planning, and interactions with outside agencies.

  6. Environment, Safety and Health Progress Assessment of the Morgantown Energy Technology Center (METC)

    SciTech Connect

    Not Available

    1993-08-01

    This report documents the result of the US Department of Energy`s (DOE) Environment, Safety and Health (ES&H) Progress Assessment of the Morgantown Energy Technology Center (METC) in Morgantown, West Virginia. METC is currently a research and development facility, managed by DOE`s Office of Fossil Energy. Its goal is to focus energy research and development to develop engineered fossil fuel systems, that are economically viable and environmentally sound, for commercial application. There is clear evidence that, since the 1991 Tiger Team Assessment, substantial progress has been made by both FE and METC in most aspects of their ES&H program. The array of new and restructured organizations, systems, and programs at FE and METC; increased assignments of staff to support these initiatives; extensive training activities; and the maturing planning processes, all reflect a discernable, continuous improvement in the quality of the ES&H performance.

  7. 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.

  8. 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.

  9. 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.

  10. Nuclear Gas Dynamics of NGC2110: A Black Hole Offset from the Host Galaxy Mass Center?

    NASA Technical Reports Server (NTRS)

    Mundell, C. G.; Ferruit, P.; Nagar, N.; Wilson, A. S.

    2004-01-01

    It has been suggested that the central regions of many galaxies are unlikely to be in a static steady state, with instabilities caused by sinking satellites, the influence of a supermassive black hole or residuals of galaxy formation, resulting in the nuclear black hole orbiting the galaxy center. The observational signature of such an orbiting black hole is an offset of the active nucleus (AGN) from the kinematic center defined by the galaxy rotation curve. This orbital motion may provide fuel for the AGN, as the hole 'grazes' on the ISM, and bent radio jets, due to the motion of their source. The early type (E/SO) Seyfert galaxy, NGC2210, with its striking twin, 'S'-shaped radio jets, is a unique and valuable test case for the offset-nucleus phenomenon since, despite its remarkably normal rotation curve, its kinematically-measured mass center is displaced both spatially (260 pc) and kinematically (170 km/s) from the active nucleus located in optical and radio studies. However, the central kinematics, where the rotation curve rises most steeply, have been inaccessible with ground-based resolutions. We present new, high resolution WFPC2 imaging and long-slit STIS spectroscopy of the central 300 pc of NGC2110. We discuss the structure and kinematics of gas moving in the galactic potential on subarcsecond scales and the reality of the offset between the black hole and the galaxy mass center.

  11. A Process-Centered Tool for Evaluating Patient Safety Performance and Guiding Strategic Improvement

    DTIC Science & Technology

    2005-01-01

    interdepartmental coordination of patient safety activities, would be crucial for introducing and improving patient safety. Two performance measures received a...seeking feedback and use of the information for improvement and creating a culture of safety. 1.2. Social Responsibility Ethical Behavior: How...all stakeholders, actively seeking feedback on patient safety and using the information for patient safety improvements . • Ensure ethical

  12. 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.

  13. Uranium Enrichment Standards of the Y-12 Nuclear Detection and Sensor Testing Center

    SciTech Connect

    Cantrell, J.

    2012-05-23

    The Y-12 National Security Complex has recently fabricated and characterized a new series of metallic uranium standards for use in the Nuclear Detection and Sensor Testing Center (NDSTC). Ten uranium metal disks with enrichments varying from 0.2 to 93.2% {sup 235}U were designed to provide researchers access to a wide variety of measurement scenarios in a single testing venue. Special care was taken in the selection of the enrichments in order to closely bracket the definitions of reactor fuel at 4% {sup 235}U and that of highly enriched uranium (HEU) at 20% {sup 235}U. Each standard is well characterized using analytical chemistry as well as a series of gamma-ray spectrometry measurements. Gamma-ray spectra of these standards are being archived in a reference library for use by customers of the NDSTC. A software database tool has been created that allows for easier access and comparison of various spectra. Information provided through the database includes: raw count data (including background spectra), regions of interest (ROIs), and full width half maximum calculations. Input is being sought from the user community on future needs including enhancements to the spectral database and additional Uranium standards, shielding configurations and detector types. A related presentation are planned for the INMM 53rd Annual Meeting (Hull, et al.), which describe new uranium chemical compound standards and testing opportunities at Y-12 Nuclear Detection and Sensor Testing Center (NDSTC).

  14. Los Alamos neutron science center nuclear weapons stewardship and unique national scientific capabilities

    SciTech Connect

    Schoenberg, Kurt F

    2010-12-15

    This presentation gives an overview of the Los Alamos Neutron Science Center (LANSCE) and its contributions to science and the nuclear weapons program. LANSCE is made of multiple experimental facilities (the Lujan Center, the Weapons Neutron Research facility (WNR), the Ultra-Cold Neutron facility (UCN), the proton Radiography facility (pRad) and the Isotope Production Facility (IPF)) served by the its kilometer long linear accelerator. Several research areas are supported, including materials and bioscience, nuclear science, materials dynamics, irradiation response and medical isotope production. LANSCE is a national user facility that supports researchers worldwide. The LANSCE Risk Mitigation program is currently in progress to update critical accelerator equipment to help extend the lifetime of LANSCE as a key user facility. The Associate Directorate of Business Sciences (ADBS) plays an important role in the continued success of LANSCE. This includes key procurement support, human resource support, technical writing support, and training support. LANSCE is also the foundation of the future signature facility MARIE (Matter-Radiation Interactions in Extremes).

  15. Improving the Safety of Oral Chemotherapy at an Academic Medical Center

    PubMed Central

    Casella, Erica; Capozzi, Donna; McGettigan, Suzanne; Gangadhar, Tara C.; Schuchter, Lynn; Myers, Jennifer S.

    2016-01-01

    Purpose: Over the last decade, the use of oral chemotherapy (OC) for the treatment of cancer has dramatically increased. Despite their route of administration, OCs pose many of the same risks as intravenous agents. In this quality improvement project, we sought to examine our current process for the prescription of OC at the Abramson Cancer Center of the University of Pennsylvania and to improve on its safety. Methods: A multidisciplinary team that included oncologists, advanced-practice providers, and pharmacists was formed to analyze the current state of our OC practice. Using Lean Six Sigma quality improvement tools, we identified a lack of pharmacist review of the OC prescription as an area for improvement. To address these deficiencies, we used our electronic medical system to route OC orders placed by treating providers to an oncology-specific outpatient pharmacist at the Abramson Cancer Center for review. Results: Over 7 months, 63 orders for OC were placed for 45 individual patients. Of the 63 orders, all were reviewed by pharmacists, and, as a result, 22 interventions were made (35%). Types of interventions included dosage adjustment (one of 22), identification of an interacting drug (nine of 22), and recommendations for additional drug monitoring (12 of 22). Conclusion: OC poses many of the same risks as intravenous chemotherapy and should be prescribed and reviewed with the same oversight. At our institution, involvement of an oncology-trained pharmacist in the review of OC led to meaningful interventions in one third of the orders. PMID:26733627

  16. 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

  17. 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

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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)

  3. Nuclear Energy Knowledge and Validation Center (NEKVaC) Needs Workshop Summary Report

    SciTech Connect

    Gougar, Hans

    2015-02-01

    The Department of Energy (DOE) has made significant progress developing simulation tools to predict the behavior of nuclear systems with greater accuracy and of increasing our capability to predict the behavior of these systems outside of the standard range of applications. These analytical tools require a more complex array of validation tests to accurately simulate the physics and multiple length and time scales. Results from modern simulations will allow experiment designers to narrow the range of conditions needed to bound system behavior and to optimize the deployment of instrumentation to limit the breadth and cost of the campaign. Modern validation, verification and uncertainty quantification (VVUQ) techniques enable analysts to extract information from experiments in a systematic manner and provide the users with a quantified uncertainty estimate. Unfortunately, the capability to perform experiments that would enable taking full advantage of the formalisms of these modern codes has progressed relatively little (with some notable exceptions in fuels and thermal-hydraulics); the majority of the experimental data available today is the "historic" data accumulated over the last decades of nuclear systems R&D. A validated code-model is a tool for users. An unvalidated code-model is useful for code developers to gain understanding, publish research results, attract funding, etc. As nuclear analysis codes have become more sophisticated, so have the measurement and validation methods and the challenges that confront them. A successful yet cost-effective validation effort requires expertise possessed only by a few, resources possessed only by the well-capitalized (or a willing collective), and a clear, well-defined objective (validating a code that is developed to satisfy the need(s) of an actual user). To that end, the Idaho National Laboratory established the Nuclear Energy Knowledge and Validation Center to address the challenges of modern code validation and to

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Safety of Dual Antiplatelet Therapy After Carotid Endarterectomy for Prevention of Restenosis: A Single Center Experience

    PubMed Central

    Barboza, Miguel A.; Chang, José; Hernández, Alvaro; Martínez, Emmanuel; Fernández, Huberth; Quirós, Gerardo; Salazar, Johanna; Ramos-Esquivel, Allan; Maud, Alberto

    2016-01-01

    Introduction The incidence of recurrent carotid stenosis after carotid endarterectomy varies from 1% to 37% with only 0–8% symptomatic restenosis. Safety of short-term (30 days) dual-antiplatelet therapy has not been established in this type of procedure. Aims To investigate the safety of dual antiplatelet therapy after carotid endarterectomy to prevent restenosis. Methods We retrospectively identified all the patients who underwent carotid endarterectomy (symptomatic or asymptomatic) treated at our center between July 2010 and July 2013 according to local protocols. All patients received a dose of 100 mg of aspirin daily immediately after carotid endarterectomy, with subsequent 100 mg of aspirin daily for the rest of the study period, and some patients received 75 mg of Clopidogrel for 30 days starting immediately after surgical procedure (dual therapy group), assigned according to medical criteria. Duplex carotid ultrasound and clinical assessments were performed at 30 days and 1 year after the procedure. Results A total of 44 patients (71.2 ± 7.9 years old; 77.2% symptomatic) were analyzed; 35 of them with dual therapy (79.54%). At 30 days, two patients from the mono-therapy group developed restenosis (22.2%), compared to none in dual therapy group (p=0.04). At one year follow-up, only one patient from the dual group showed restenosis (p=0.10). No deaths, major bleeding or new strokes were reported in both groups. Conclusions Short-term dual antiplatelet therapy with aspirin and clopidogrel after carotid endarterectomy might be associated with a lower incidence of restenosis. This observation must be validated in a prospective trial. PMID:27829964

  13. 77 FR 44544 - Safety Zone; Gilmerton Bridge Center Span Float-In, Elizabeth River; Norfolk, Portsmouth, and...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-07-30

    ... SECURITY Coast Guard 33 CFR Part 165 RIN 1625-AA00 Safety Zone; Gilmerton Bridge Center Span Float-In... Span Float-in and bridge construction of span placement. The original proposal had a start date of July... span lift construction. DATES: The proposed rule is withdrawn on July 6, 2012. ADDRESSES: The...

  14. 77 FR 75016 - Safety Zone: Gilmerton Bridge Center Span Float-in, Elizabeth River; Norfolk, Portsmouth, and...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-19

    ... center span barge. Inadvertently, this rule included an error in the inclement weather date of the safety... January 11, 2013, with inclement weather dates of January 12, 2013 through January 16, 2013. ADDRESSES... 11, 2013, with inclement weather dates of January 12, 2013 through January 16, 2013. However, due...

  15. Commercialization of Kennedy Space Center Instrumentation Developed to Improve Safety, Reliability, Cost Effectiveness of Space Shuttle Processing, Launch, and Landing

    NASA Technical Reports Server (NTRS)

    Helms, William R.; Starr, Stanley O.

    1997-01-01

    Priorities and achievements of the Kennedy Space Center (KSF) Instrumentation Laboratories in improving operational safety and decreasing processing costs associated with the Shuttle vehicle are addressed. Technologies that have been or are in the process of technology transfer are reviewed, and routes by which commercial concerns can obtain licenses to other KSF Instrumentation Laboratory technologies are discussed.

  16. Food Safety: at the center of a One Health approach for combating zoonoses.

    PubMed

    Wielinga, Peter R; Schlundt, Jørgen

    2013-01-01

    Food Safety is at the center of One Health. Many, if not most, of all important zoonoses relate in some way to animals in the food production chain. Therefore, the food becomes an important vehicle for many, but not all, of these zoonotic pathogens. One of the major issues in food safety over the latest decennia has been the lack of cross-sectoral collaboration across the food production chain. Major food safety events have been significantly affected by the lack of collaboration between the animal health, the food control, and the human health sector. Examples range from BSE and E. coli outbreaks over dioxin crises to intentional melamine contamination. One Health formulates clearly both the need for and the benefit of cross-sectoral collaboration. In this chapter, we will focus on the human health risk related to zoonotic microorganisms present both in food animals and food from these animals, and typically transmitted to humans through food. We focus on these issues because they are very important in relation to the human disease burden, but also because this is the area where some experience of cross-sectoral collaboration already exist. Food related zoonoses can be separated in three major classes: parasites, bacteria, and viruses. While parasites often relate to very specific animal hosts and contribute significantly to the human disease burden, virus have often been related to major, well-published global outbreaks, e.g. SARS and avian- and swine-influenza. The bacterial zoonoses on the other hand often result in sporadic, but very wide-spread disease cases, resulting in a major disease burden in all countries, e.g. Salmonella and Campylobacter. Next to these traditional zoonotic problems, the use of antimicrobials in (food) animals has also caused the emergence of antimicrobial resistant (AMR) zoonotic bacteria. It is important to realize the difference in the nature of disease epidemiology, as well as, in society's reaction to these diseases in different

  17. 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.

  18. Cooperative Monitoring Center Occasional Paper/6: Pakistani Perceptions and Prospects of Reducing the Nuclear Danger in South Asia

    SciTech Connect

    Kamal, N.

    1999-01-01

    The Indian and Pakistani nuclear tests in May 1998 triggered a full-blown nuclear debate. For the first time, hard-liners, moderates, and pacifists engaged in an extensive public discussion that helped to make the people of Pakistan more sensitive to the dangers of nuclear competition. Pakistan's concerns about its conventional military inferiority, both in the present and future, and the belief that nuclear capability would deter India from exerting its superior military strength, constituted the bedrock of its perception on the nuclear issue. Ofilcial Pakistani statements, both immediately after the nuclear tests and later, have advocated restraint on the issue of nuclearization, indicating cognizance of the importance of avoiding a regional nuclear arms competition, both for security and economic reasons. This paper suggests a variety of nonweaponization and nondeployment options that would serve the security interests of India and Pakistan. Besides preventing a hair-trigger situation, these options could reduce the financial and logistical burden of ensuring the safety and security of nuclear weapons as well as lower strategic threat-perceptions.

  19. The Rhode Island Nuclear Science Center conversion from HEU to LEU fuel

    SciTech Connect

    Tehan, Terry

    2000-09-27

    The 2-MW Rhode Island Nuclear Science Center (RINSC) open pool reactor was converted from 93% UAL-High Enriched Uranium (HEU) fuel to 20% enrichment U3Si2-AL Low Enriched Uranium (LEU) fuel. The conversion included redesign of the core to a more compact size and the addition of beryllium reflectors and a beryllium flux trap. A significant increase in thermal flux level was achieved due to greater neutron leakage in the new compact core configuration. Following the conversion, a second cooling loop and an emergency core cooling system were installed to permit operation at 5 MW. After re-licensing at 2 MW, a power upgrade request will be submitted to the NRC.

  20. Nuclear counting filter based on a centered Skellam test and a double exponential smoothing

    SciTech Connect

    Coulon, Romain; Kondrasovs, Vladimir; Dumazert, Jonathan; Rohee, Emmanuel; Normand Stephane

    2015-07-01

    Online nuclear counting represents a challenge due to the stochastic nature of radioactivity. The count data have to be filtered in order to provide a precise and accurate estimation of the count rate, this with a response time compatible with the application in view. An innovative filter is presented in this paper addressing this issue. It is a nonlinear filter based on a Centered Skellam Test (CST) giving a local maximum likelihood estimation of the signal based on a Poisson distribution assumption. This nonlinear approach allows to smooth the counting signal while maintaining a fast response when brutal change activity occur. The filter has been improved by the implementation of a Brown's double Exponential Smoothing (BES). The filter has been validated and compared to other state of the art smoothing filters. The CST-BES filter shows a significant improvement compared to all tested smoothing filters. (authors)

  1. Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center

    SciTech Connect

    M. D. Staiger

    1999-06-01

    A potential option in the program for long-term management of high-level wastes at the Idaho Nuclear Technology and Engineering Center (INTEC), at the Idaho National Engineering and Environmental Laboratory, calls for retrieving calcine waste and converting it to a more stable and less dispersible form. An inventory of calcine produced during the period December 1963 to May 1999 has been prepared based on calciner run, solids storage facilities operating, and miscellaneous operational information, which gives the range of chemical compositions of calcine waste stored at INTEC. Information researched includes calciner startup data, waste solution analyses and volumes calcined, calciner operating schedules, solids storage bin capacities, calcine storage bin distributor systems, and solids storage bin design and temperature monitoring records. Unique information on calcine solids storage facilities design of potential interest to remote retrieval operators is given.

  2. Aging Management Guideline for commercial nuclear power plants: Motor control centers; Final report

    SciTech Connect

    Toman, G.; Gazdzinski, R.; O`Hearn, E.

    1994-02-01

    This Aging Management Guideline (AMG) provides recommended methods for effective detection and mitigation of age-related degradation mechanisms in Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) commercial nuclear power plant motor control centers important to license renewal. The intent of this AMG is to assist plant maintenance and operations personnel in maximizing the safe, useful life of these components. It also supports the documentation of effective aging management programs required under the License Renewal Rule 10 CFR Part 54. This AMG is presented in a manner that allows personnel responsible for performance analysis and maintenance to compare their plant-specific aging mechanisms (expected or already experienced) and aging management program activities to the more generic results and recommendations presented herein.

  3. Executive summary. Conceptual studies nuclear energy center Lake Hartwell, S. C. , Phase III

    SciTech Connect

    Not Available

    1981-01-01

    This document summarizes a conceptual study on the feasibility and practicality of developing a nuclear energy center (NEC) at a specific site in the SSEB region. The site selected for this conceptual study is at Lake Hartwell, South Carolina. The conceptual NEC at Lake Hartwell consists of twelve 1250-MW(e) LWRs arranged on the site in four cluster of three units each, know as triads. The nominal distance between triads was selected as 2-1/2 miles. The total electric output of 15,000 MWe to be generated by the NEC would be transmitted to five major utilities in South Carolina, North Carolina, and Georgia. Objective of the study was to assess the technical, socioeconomic, environmental, and institutional issues relating to the NEC at the conceptual study site. The basic finding was that the concept of a NEC on the Lake Hartwell site is feasible, but further analysis of institutional issues and possible legislation would be required.

  4. THE OFF-CENTERED SEYFERT-LIKE COMPACT EMISSION IN THE NUCLEAR REGION OF NGC 3621

    SciTech Connect

    Menezes, R. B.; Steiner, J. E.; Silva, Patricia da

    2016-02-01

    We analyze an optical data cube of the nuclear region of NGC 3621, taken with the integral field unit of the Gemini Multi-object Spectrograph. We found that the previously detected central line emission in this galaxy actually comes from a blob, located at a projected distance of 2.″14 ± 0.″08 (70.1 ± 2.6 pc) from the stellar nucleus. Only diffuse emission was detected in the rest of the field of view, with a deficit of emission at the position of the stellar nucleus. Diagnostic diagram analysis reveals that the off-centered emitting blob has a Seyfert 2 spectrum. We propose that the line-emitting blob may be a “fossil” emission-line region or a light “echo” from an active galactic nucleus (AGN), which was significantly brighter in the past. Our estimates indicate that the bolometric luminosity of the AGN must have decreased by a factor of ∼13–500 during the past ∼230 yr. A second scenario to explain the morphology of the line-emitting areas in the nuclear region of NGC 3621 involves no decrease of the AGN bolometric luminosity and establishes that the AGN is highly obscured toward the observer but not toward the line-emitting blob. The third scenario proposed here assumes that the off-centered line-emitting blob is a recoiling supermassive black hole, after the coalescence of two black holes. Finally, an additional hypothesis is that the central X-ray source is not an AGN, but an X-ray binary. This idea is consistent with all the scenarios we proposed.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  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

    1986-07-01

    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.

  10. 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.

  11. 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.

  12. Safety Analysis Report for Packaging (SARP): Model AL-M1 nuclear packaging (DOE C of C No. USA/9507/BLF)

    SciTech Connect

    Coleman, H.L.; Whitney, M.A.; Williams, M.A.; Alexander, B.M.; Shapiro, A.

    1987-11-24

    This Safety Analysis Report for Packaging (SARP) satisfies the request of the US Department of Energy for a formal safety analysis of the shipping container identified as USA/9507/BLF, also called AL-M1, configuration 5. This report makes available to all potential users the technical information and the limits pertinent to the construction and use of the shipping containers. It includes discussions of structural integrity, thermal resistance, radiation shielding and radiological safety, nuclear criticality safety, and quality control. A complete physical and technical description of the package is presented. The package consists of an inner container centered within an insulated steel drum. The configuration-5 package contains tritiated water held on sorbent material. There are two other AL-M1 packages, designated configurations 1 and 3. These use the same insulated outer drum, but licensing of these containers will not be addressed in this SARP. Design and development considerations, the tests and evaluations required to prove the ability of the container to withstand normal transportation conditions, and the sequence of four hypothetical accident conditions (free drop, puncture, thermal, and water immersion) are discussed. Tables, graphs, dimensional sketches, photographs, technical references, loading and shipping procedures, Monsanto Research Corporation-Mound experience in using the containers, and a copy of the DOE/OSD/ALO Certificate of Compliance are included.

  13. NASA Engineering Safety Center NASA Aerospace Flight Battery Systems Working Group 2007 Proactive Task Status

    NASA Technical Reports Server (NTRS)

    Manzo, Michelle A.

    2007-01-01

    In 2007, the NASA Engineering Safety Center (NESC) chartered the NASA Aerospace Flight Battery Systems Working Group to bring forth and address critical battery-related performance/manufacturing issues for NASA and the aerospace community. A suite of tasks identifying and addressing issues related to Ni-H2 and Li-ion battery chemistries was submitted and selected for implementation. The current NESC funded are: (1) Wet Life of Ni-H2 Batteries (2) Binding Procurement (3) NASA Lithium-Ion Battery Guidelines (3a) Li-Ion Performance Assessment (3b) Li-Ion Guidelines Document (3b-i) Assessment of Applicability of Pouch Cells for Aerospace Missions (3b-ii) High Voltage Risk Assessment (3b-iii) Safe Charge Rates for Li-Ion Cells (4) Availability of Source Material for Li-Ion Cells (5) NASA Aerospace Battery Workshop This presentation provides a brief overview of the tasks in the 2007 plan and serves as an introduction to more detailed discussions on each of the specific tasks.

  14. Materials characterization center workshop on the irradiation effects in nuclear waste forms

    SciTech Connect

    Roberts, F.P.; Turcotte, R.P.; Weber, W.J.

    1981-01-01

    The Workshop on Irradiation Effects in Nuclear Waste Forms sponsored by the Materials Characterization Center (MCC) brought together experts in radiation damage in materials and waste-management technology to review the problems associated with irradiation effects on waste-form integrity and to evaluate standard methods for generating data to be included in the Nuclear Waste Materials Handbook. The workshop reached the following conclusions: the concept of Standard Test for the Effects of Alpha-Decay in Nuclear Waste Solids, (MCC-6) for evaluating the effects of alpha decay is valid and useful, and as a result of the workshop, modifications to the proposed procedure will be incorpoated in a revised version of MCC-6; the MCC-6 test is not applicable to the evaluation of radiation damage in spent fuel; plutonium-238 is recommended as the dopant for transuranic and defense high-level waste forms, and when high doses are required, as in the case of commercial high-level waste forms, /sup 244/Cm can be used; among the important property changes caused by irradiation are those that lead to greater leachability, and additionally, radiolysis of the leachant may increase leach rates; research is needed in this area; ionization-induced changes in physical properties can be as important as displacement damage in some materials, and a synergism is also likely to exist from the combined effects of ionization and displacement damage; and the effect of changing the temperature and dose rates on property changes induced by radiation damage needs to be determined.

  15. 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.

  16. 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.

  17. 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

  18. A National Tracking Center for Monitoring Shipments of HEU, MOX, and Spent Nuclear Fuel: How do we implement?

    SciTech Connect

    Mark Schanfein

    2009-07-01

    Nuclear material safeguards specialists and instrument developers at US Department of Energy (USDOE) National Laboratories in the United States, sponsored by the National Nuclear Security Administration (NNSA) Office of NA-24, have been developing devices to monitor shipments of UF6 cylinders and other radioactive materials , . Tracking devices are being developed that are capable of monitoring shipments of valuable radioactive materials in real time, using the Global Positioning System (GPS). We envision that such devices will be extremely useful, if not essential, for monitoring the shipment of these important cargoes of nuclear material, including highly-enriched uranium (HEU), mixed plutonium/uranium oxide (MOX), spent nuclear fuel, and, potentially, other large radioactive sources. To ensure nuclear material security and safeguards, it is extremely important to track these materials because they contain so-called “direct-use material” which is material that if diverted and processed could potentially be used to develop clandestine nuclear weapons . Large sources could be used for a dirty bomb also known as a radioactive dispersal device (RDD). For that matter, any interdiction by an adversary regardless of intent demands a rapid response. To make the fullest use of such tracking devices, we propose a National Tracking Center. This paper describes what the attributes of such a center would be and how it could ultimately be the prototype for an International Tracking Center, possibly to be based in Vienna, at the International Atomic Energy Agency (IAEA).

  19. Dynamic two-center interference in high-order harmonic generation from molecules with attosecond nuclear motion.

    PubMed

    Baker, S; Robinson, J S; Lein, M; Chirilă, C C; Torres, R; Bandulet, H C; Comtois, D; Kieffer, J C; Villeneuve, D M; Tisch, J W G; Marangos, J P

    2008-08-01

    We report a new dynamic two-center interference effect in high-harmonic generation from H2, in which the attosecond nuclear motion of H2+ initiated at ionization causes interference to be observed at lower harmonic orders than would be the case for static nuclei. To enable this measurement we utilize a recently developed technique for probing the attosecond nuclear dynamics of small molecules. The experimental results are reproduced by a theoretical analysis based upon the strong-field approximation which incorporates the temporally dependent two-center interference term.

  20. Sensing of single nuclear spins in random thermal motion with proximate nitrogen-vacancy centers

    NASA Astrophysics Data System (ADS)

    Bruderer, M.; Fernández-Acebal, P.; Aurich, R.; Plenio, M. B.

    2016-03-01

    Nitrogen-vacancy (NV) centers in diamond have emerged as valuable tools for sensing and polarizing spins. Motivated by potential applications in chemistry, biology, and medicine, we show that NV-based sensors are capable of detecting single spin targets even if they undergo diffusive motion in an ambient thermal environment. Focusing on experimentally relevant diffusion regimes, we derive an effective model for the NV-target interaction, where parameters entering the model are obtained from numerical simulations of the target motion. The practicality of our approach is demonstrated by analyzing two realistic experimental scenarios: (i) time-resolved sensing of a fluorine nuclear spin bound to an N-heterocyclic carbene-ruthenium (NHC-Ru) catalyst that is immobilized on the diamond surface and (ii) detection of an electron spin label by an NV center in a nanodiamond, both attached to a vibrating chemokine receptor in thermal motion. We find in particular that the detachment of a fluorine target from the NHC-Ru carrier molecule can be monitored with a time resolution of a few seconds.

  1. 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.

  2. 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

    ... Public and the Workers, to the Department of Energy. In accordance with section 315(b) of the Atomic... approving Documented Safety Analyses for nuclear facilities. The Recommendation identified six specific sub... provided clarifications regarding the purposes for each sub- recommendation and stated that there...

  3. 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

  4. 21 CFR Appendix A to Part 101 - Monier-Williams Procedure (With Modifications) for Sulfites in Food, Center for Food Safety and...

    Code of Federal Regulations, 2014 CFR

    2014-04-01

    ... 21 Food and Drugs 2 2014-04-01 2014-04-01 false Monier-Williams Procedure (With Modifications) for Sulfites in Food, Center for Food Safety and Applied Nutrition, Food and Drug Administration (November 1985...-Williams Procedure (With Modifications) for Sulfites in Food, Center for Food Safety and Applied...

  5. 21 CFR Appendix A to Part 101 - Monier-Williams Procedure (With Modifications) for Sulfites in Food, Center for Food Safety and...

    Code of Federal Regulations, 2010 CFR

    2010-04-01

    ... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Monier-Williams Procedure (With Modifications) for Sulfites in Food, Center for Food Safety and Applied Nutrition, Food and Drug Administration (November 1985...-Williams Procedure (With Modifications) for Sulfites in Food, Center for Food Safety and Applied...

  6. Computer-generated formulas for three-center nuclear-attraction integrals (electrostatic potential) for Slater-type orbitals

    NASA Technical Reports Server (NTRS)

    Jones, H. W.

    1984-01-01

    The computer-assisted C-matrix, Loewdin-alpha-function, single-center expansion method in spherical harmonics has been applied to the three-center nuclear-attraction integral (potential due to the product of separated Slater-type orbitals). Exact formulas are produced for 13 terms of an infinite series that permits evaluation to ten decimal digits of an example using 1s orbitals.

  7. 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.

  8. TWRS safety program plan

    SciTech Connect

    Calderon, L.M., Westinghouse Hanford

    1996-08-01

    Management of Nuclear Safety, Industrial Safety, Industrial Hygiene, and Fire Protection programs, functions, and field support resources for Tank Waste Remediation Systems (TWRS) has, until recently, been centralized in TWRS Safety, under the Emergency, Safety, and Quality organization. Industrial hygiene technician services were also provided to support operational needs related to safety basis compliance. Due to WHC decentralization of safety and reengineering efforts in West Tank Farms, staffing and safety responsibilities have been transferred to the facilities. Under the new structure, safety personnel for TWRS are assigned directly to East Tank Farms, West Tank Farms, and a core Safety Group in TWRS Engineering. The Characterization Project Operations (CPO) safety organization will remain in tact as it currently exists. Personnel assigned to East Tank Farms, West Tank Farms, and CPO will perform facility-specific or project-specific duties and provide field implementation of programs. Those assigned to the core group will focus on activities having a TWRS-wide or programmatic focus. Hanford-wide activities will be the responsibility of the Safety Center of Expertise. In order to ensure an effective and consistent safety program for TWRS under the new organization program functions, goals, organizational structure, roles, responsibilities, and path forward must be clearly established. The purpose of the TWRS Safety Program Plan is to define the overall safety program, responsibilities, relationships, and communication linkages for safety personnel under the new structure. In addition, issues associated with reorganization transition are addressed, including training, project ownership, records management, and dissemination of equipment. For the purpose of this document ``TWRS Safety`` refers to all safety professionals and technicians (Industrial Safety, Industrial Hygiene, Fire Protection, and Nuclear Safety) within the TWRS organization, regardless of their

  9. MRI with ferumoxytol: A single center experience of safety across the age spectrum

    PubMed Central

    Nguyen, Kim‐Lien; Yoshida, Takegawa; Han, Fei; Ayad, Ihab; Reemtsen, Brian L.; Salusky, Isidro B.; Satou, Gary M.; Hu, Peng

    2016-01-01

    Purpose To summarize our single‐center safety experience with the off‐label use of ferumoxytol for magnetic resonance imaging (MRI) and to compare the effects of ferumoxytol on monitored physiologic indices in patients under anesthesia with those of gadofosveset trisodium. Materials and Methods Consecutive patients who underwent ferumoxytol‐enhanced (FE) MRI exams were included. Adverse events (AEs) were classified according to the Common Terminology Criteria for Adverse Events v4.0. In a subgroup of patients examined under general anesthesia, recording of blood pressure, heart rate, oxygen saturation, and end‐tidal CO2 was performed. A comparable group of 23 patients who underwent gadofosveset‐enhanced (GE) MRI under anesthesia with similar monitoring was also analyzed. Results In all, 217 unique patients, ages 3 days to 94 years, underwent FE‐MRI. No ferumoxytol‐related severe, life‐threatening, or fatal AEs occurred acutely or at follow‐up. Two patients developed ferumoxytol‐related nausea. Between‐group (FE‐ vs. GE‐MRI) comparisons showed no statistical difference in heart rate (P = 0.69, 95% confidence interval [CI] 96–113 bpm), mean arterial blood pressure (MAP) (P = 0.74, 95% CI 44–52 mmHg), oxygen saturation (P = 0.76, 95% CI 94–98%), and end‐tidal CO2 (P = 0.73, 95% CI 31–37 mmHg). No significant change in MAP (P = 0.12, 95% CI 50–58 mmHg) or heart rate (P = 0.25, 95% CI 91–105 bpm) was noted between slow infusion of ferumoxytol (n = 113) vs. bolus injection (n = 104). Conclusion In our single‐center experience, no serious AEs occurred with the diagnostic use of ferumoxytol across a wide spectrum of age, renal function, and indications. Because of the limited sample size, firm conclusions cannot be drawn about the generalizability of our results. Thus, vigilance and monitoring are recommended to mitigate potential rare adverse reactions. Level of Evidence: 2 J. Magn. Reson. Imaging

  10. Peripherally inserted central venous catheter safety in burn care: a single-center retrospective cohort review.

    PubMed

    Austin, Ryan E; Shahrokhi, Shahriar; Bolourani, Siavash; Jeschke, Marc G

    2015-01-01

    The use of peripherally inserted central catheter (PICC) line for central venous access in thermally injured patients has increased in recent years despite a lack of evidence regarding safety in this patient population. A recent survey of invasive catheter practices among 44 burn centers in the United States found that 37% of burn units use PICC lines as part of their treatment protocol. The goal of this study was to compare PICC-associated complication rates with the existing literature in both the critical care and burn settings. The methodology involved is a single institution retrospective cohort review of patients who received a PICC line during admission to a regional burn unit between 2008 and 2013. Fifty-three patients were identified with a total of seventy-three PICC lines. The primary outcome measurement for this study was indication for PICC line discontinuation. The most common reason for PICC line discontinuation was that the line was no longer indicated (45.2%). Four cases of symptomatic upper extremity deep vein thrombosis (5.5%) and three cases of central line-associated bloodstream infection (4.3%, 2.72 infections per 1000 line days) were identified. PICC lines were in situ an average of 15 days (range 1 to 49 days). We suggest that PICC line-associated complication rates are similar to those published in the critical care literature. Though these rates are higher than those published in the burn literature, they are similar to central venous catheter-associated complication rates. While PICC lines can be a useful resource in the treatment of the thermally injured patient, they are associated with significant and potentially fatal risks.

  11. Theoretical studies in medium-energy nuclear and hadronic physics. [Indiana Univ. Nuclear Theory Center and Department of Physics

    SciTech Connect

    Horowitz, C J; Macfarlane, M H; Matsui, T; Serot, B D

    1993-01-01

    A proposal for theoretical nuclear physics research is made for the period April 1, 1993 through March 31, 1996. Research is proposed in the following areas: relativistic many-body theory of nuclei and nuclear matter, quasifree electroweak scattering and strange quarks in nuclei, dynamical effects in (e,e[prime]p) scattering at large momentum transfer, investigating the nucleon's parton sea with polarized leptoproduction, physics of ultrarelativistic nucleus[endash]nucleus collisions, QCD sum rules and hadronic properties, non-relativistic models of nuclear reactions, and spin and color correlations in a quark-exchange model of nuclear matter. Highlights of recent research, vitae of principal investigators, and lists of publications and invited talks are also given. Recent research dealt primarily with medium-energy nuclear physics, relativistic theories of nuclei and the nuclear response, the nuclear equation of state under extreme conditions, the dynamics of the quark[endash]gluon plasma in relativistic heavy-ion collisions, and theories of the nucleon[endash]nucleon force.

  12. 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

  13. 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).

  14. 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).

  15. 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)

  16. 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.

  17. 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.

  18. [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.

  19. 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.

  20. Report of a workshop on nuclear forces and nonproliferation Woodrow Wilson international center for scholars, Washington, DC October 28, 2010

    SciTech Connect

    Pilat, Joseph F

    2010-12-08

    A workshop sponsored by the Los Alamos National Laboratory in cooperation with the Woodrow Wilson International Center for Scholars was held at the Wilson Center in Washington, DC, on October 28, 2010. The workshop addressed evolving nuclear forces and their impacts on nonproliferation in the context of the new strategic environment, the Obama Administration's Nuclear Posture Review and the 2010 NPT Review Conference. The discussions reflected the importance of the NPR for defining the role of US nuclear forces in dealing with 21st century threats and providing guidance for National Nuclear Security Administration (NNSA) and Department of Defense (DoD) programs and, for many but not all participants, highlighted its role in the successful outcome of the NPT RevCon. There was widespread support for the NPR and its role in developing the foundations for a sustainable nuclear-weapon program that addresses nuclear weapons, infrastructure and expertise in the broader nonproliferation, disarmament and international security contexts. However, some participants raised concerns about its implementation and its long-term effectiveness and sustainability.