Science.gov

Sample records for nuclear safety element

  1. 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.

  2. 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…

  3. 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.

  4. Use of molybdenum as a structural material of fuel elements for improving the safety of nuclear reactors

    NASA Astrophysics Data System (ADS)

    Shmelev, A. N.; Kozhahmet, B. K.

    2017-01-01

    Main purpose of the study is justifying the use of molybdenum as a structural material of fuel elements for improving the safety of nuclear reactors. Particularity of used molybdenum is that its isotopic composition corresponds to molybdenum, which is obtained as the tailing during operation of the separation cascade for producing a material for medical diagnostics of cancer. When performing the study the neutron-physical properties of isotopes of natural molybdenum (nuclear data library JENDL-4.0) and thermal properties of metallic molybdenum were used. The following results were obtained: 1. A method for reducing the thermal constant of fuel elements for light water and fast reactors by using dispersion fuel in cylindrical fuel rods containing, for example, granules of metallic U-Mo-alloy into Mo-matrix was proposed. 2. The necessity of molybdenum enrichment by weakly absorbing isotopes was shown. 3. Total use of isotopic molybdenum will be more than 50%. A method for reducing the thermal constant of the fuel elements, allowing us to increase the safety of light water and fast nuclear reactors by using dispersion fuel in cylindrical fuel rods containing, for example, granules of metallic U-Mo-alloy into Mo-matrix with enrichment by weakly absorbing isotopes of molybdenum is proposed.

  5. Nuclear fuel element

    DOEpatents

    Meadowcroft, Ronald Ross; Bain, Alastair Stewart

    1977-01-01

    A nuclear fuel element wherein a tubular cladding of zirconium or a zirconium alloy has a fission gas plenum chamber which is held against collapse by the loops of a spacer in the form of a tube which has been deformed inwardly at three equally spaced, circumferential positions to provide three loops. A heat resistant disc of, say, graphite separates nuclear fuel pellets within the cladding from the plenum chamber. The spacer is of zirconium or a zirconium alloy.

  6. 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.

  7. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1980-04-29

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

  8. Nuclear fuel element

    DOEpatents

    Armijo, Joseph S.; Coffin, Jr., Louis F.

    1983-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has a composite cladding having a substrate and a metal barrier metallurgically bonded on the inside surface of the substrate so that the metal barrier forms a shield between the substrate and the nuclear fuel material held within the cladding. The metal barrier forms about 1 to about 30 percent of the thickness of the cladding and is comprised of a low neutron absorption metal of substantially pure zirconium. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the substrate from contact and reaction with such impurities and fission products. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy. Methods of manufacturing the composite cladding are also disclosed.

  9. Nuclear fuel element

    DOEpatents

    Zocher, Roy W.

    1991-01-01

    A nuclear fuel element and a method of manufacturing the element. The fuel element is comprised of a metal primary container and a fuel pellet which is located inside it and which is often fragmented. The primary container is subjected to elevated pressure and temperature to deform the container such that the container conforms to the fuel pellet, that is, such that the container is in substantial contact with the surface of the pellet. This conformance eliminates clearances which permit rubbing together of fuel pellet fragments and rubbing of fuel pellet fragments against the container, thus reducing the amount of dust inside the fuel container and the amount of dust which may escape in the event of container breach. Also, as a result of the inventive method, fuel pellet fragments tend to adhere to one another to form a coherent non-fragmented mass; this reduces the tendency of a fragment to pierce the container in the event of impact.

  10. Nuclear reactor safety device

    DOEpatents

    Hutter, Ernest

    1986-01-01

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

  11. Nuclear reactor fuel element

    DOEpatents

    Johnson, Carl E.; Crouthamel, Carl E.

    1980-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of oxygen gettering material on the inner surface of the cladding. The gettering material reacts with oxygen released by the fissionable material during irradiation of the core thereby preventing the oxygen from reacting with and corroding the cladding. Also described is an improved method for coating the inner surface of the cladding with a layer of gettering material.

  12. 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.

  13. Nuclear reactor safety device

    DOEpatents

    Hutter, E.

    1983-08-15

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

  14. 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…

  15. 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.

  16. Vented nuclear fuel element

    DOEpatents

    Grossman, Leonard N.; Kaznoff, Alexis I.

    1979-01-01

    A nuclear fuel cell for use in a thermionic nuclear reactor in which a small conduit extends from the outside surface of the emitter to the center of the fuel mass of the emitter body to permit escape of volatile and gaseous fission products collected in the center thereof by virtue of molecular migration of the gases to the hotter region of the fuel.

  17. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  18. NUCLEAR REACTOR FUEL ELEMENT

    DOEpatents

    Anderson, W.F.; Tellefson, D.R.; Shimazaki, T.T.

    1962-04-10

    A plate type fuel element which is particularly useful for organic cooled reactors is described. Generally, the fuel element comprises a plurality of fissionable fuel bearing plates held in spaced relationship by a frame in which the plates are slidably mounted in grooves. Clearance is provided in the grooves to allow the plates to expand laterally. The plates may be rigidly interconnected but are floatingly supported at their ends within the frame to allow for longi-tudinal expansion. Thus, this fuel element is able to withstand large temperature differentials without great structural stresses. (AEC)

  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. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-16

    A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

  1. NUCLEAR REACTOR ELEMENT

    DOEpatents

    Sanz, M.C.; Scully, C.N.

    1961-06-27

    The patented fuel element is a hexagonal graphite body having an axial channel therethrough. The graphite is impregnated with uranium which is concentrated near the axial channel. Layers of tantalum nitride and tantalum carbide are disposed on the surface of the body confronting the channel.

  2. 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.

  3. 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...

  4. 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}.

  5. NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY

    DOEpatents

    Stengel, F.G.

    1963-12-24

    A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

  6. 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.

  7. 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...

  8. 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...

  9. 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...

  10. 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...

  11. 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.

  12. 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.

  13. Seven Basic Elements of a Safety Program.

    ERIC Educational Resources Information Center

    Oliphant, Richard J.

    1979-01-01

    Outlined are the basic elements of a strong utility employee safety program. The components discussed include: management leadership; assignment of responsibility; maintenance; establishment of safety training; accident record system; medical systems; and personal responsibility of employees. (CS)

  14. 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.

  15. 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.

  16. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-05-01

    A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

  17. 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.

  18. 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…

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. CONSTRUCTION OF NUCLEAR FUEL ELEMENTS

    DOEpatents

    Weems, S.J.

    1963-09-24

    >A rib arrangement and an end construction for nuclearfuel elements laid end to end in a coolant tube are described. The rib arrangement is such that each fuel element, when separated from other fuel elements, fits loosely in the coolant tube and so can easily be inserted or withdrawn from the tube. The end construction of the fuel elements is such that the fuel elements when assembled end to end are keyed against relative rotation, and the ribs of each fuel element cooperate with the ribs of the adjacent fuel elements to give the assembled fuel elements a tight fit with the coolant tube. (AEC)

  5. 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

  6. 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…

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. Monitoring arrangement for vented nuclear fuel elements

    DOEpatents

    Campana, Robert J.

    1981-01-01

    In a nuclear fuel reactor core, fuel elements are arranged in a closely packed hexagonal configuration, each fuel element having diametrically opposed vents permitting 180.degree. rotation of the fuel elements to counteract bowing. A grid plate engages the fuel elements and forms passages for communicating sets of three, four or six individual vents with respective monitor lines in order to communicate vented radioactive gases from the fuel elements to suitable monitor means in a manner readily permitting detection of leakage in individual fuel elements.

  15. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Dickson, J.J.

    1963-09-24

    A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

  16. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    DOEpatents

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  17. Damper mechanism for nuclear reactor control elements

    DOEpatents

    Taft, William Elwood

    1976-01-01

    A damper mechanism which provides a nuclear reactor control element decelerating function at the end of the scram stroke. The total damping function is produced by the combination of two assemblies, which operate in sequence. First, a tapered dashram assembly decelerates the control element to a lower velocity, after which a spring hydraulic damper assembly takes over to complete the final damping.

  18. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

  19. Rack for storing spent nuclear fuel elements

    DOEpatents

    Rubinstein, Herbert J.; Clark, Philip M.; Gilcrest, James D.

    1978-06-20

    A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed fuel elements. The enclosures are fixed at the lower ends thereof to a base. Pockets are formed between confronting walls of adjacent enclosures for receiving high absorption neutron absorbers, such as Boral, cadmium, borated stainless steel and the like for the closer spacing of spent fuel elements.

  20. FOIL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Noland, R.A.; Walker, D.E.; Spinrad, B.I.

    1963-07-16

    A method of making a foil-type fuel element is described. A foil of fuel metal is perforated in; regular design and sheets of cladding metal are placed on both sides. The cladding metal sheets are then spot-welded to each other through the perforations, and the edges sealed. (AEC)

  1. FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Bassett, C.H.

    1961-11-21

    A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

  2. Nuclear fuel elements having a composite cladding

    DOEpatents

    Gordon, Gerald M.; Cowan, II, Robert L.; Davies, John H.

    1983-09-20

    An improved nuclear fuel element is disclosed for use in the core of nuclear reactors. The improved nuclear fuel element has a composite cladding of an outer portion forming a substrate having on the inside surface a metal layer selected from the group consisting of copper, nickel, iron and alloys of the foregoing with a gap between the composite cladding and the core of nuclear fuel. The nuclear fuel element comprises a container of the elongated composite cladding, a central core of a body of nuclear fuel material disposed in and partially filling the container and forming an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container and a nuclear fuel material retaining means positioned in the cavity. The metal layer of the composite cladding prevents perforations or failures in the cladding substrate from stress corrosion cracking or from fuel pellet-cladding interaction or both. The substrate of the composite cladding is selected from conventional cladding materials and preferably is a zirconium alloy.

  3. 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.

  4. 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.

  5. Management of National Nuclear Power Programs for assured safety

    SciTech Connect

    Connolly, T.J.

    1985-01-01

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

  6. NUCLEAR SAFETY 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.

  7. 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.

  8. Low exchange element for nuclear reactor

    DOEpatents

    Brogli, Rudolf H.; Shamasunder, Bangalore I.; Seth, Shivaji S.

    1985-01-01

    A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

  9. 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.

  10. 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

  11. 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.

  12. 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)

  13. Frontiers of heavy element nuclear and radiochemistry

    SciTech Connect

    Hoffman, D.C.

    1997-10-01

    The production and half-lives of the heaviest chemical elements, now known through Z = 112, are reviewed. Recent experimental evidence for the stabilization of heavy element isotopes due to proximity to deformed nuclear shells at Z = 108 and N = 162 is compared with the theoretical predictions. The possible existence of isotopes of elements 107--110 with half-lives of seconds or longer, and production reactions and experimental techniques for increasing the overall yields of such isotopes in order to study both their nuclear and chemical properties are discussed. The present status of studies of the chemical properties of Rf, Ha, and Sg is briefly summarized and prospects for extending chemical studies beyond Sg are considered.

  14. 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.

  15. 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)

  16. Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Astrophysics Data System (ADS)

    Emrich, William J.

    2008-01-01

    To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

  17. Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    SciTech Connect

    Emrich, William J. Jr.

    2008-01-21

    To support a potential future development of a nuclear thermal rocket engine, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components could be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

  18. 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.

  19. 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.

  20. Single cell elemental analysis using nuclear microscopy

    NASA Astrophysics Data System (ADS)

    Ren, M. Q.; Thong, P. S. P.; Kara, U.; Watt, F.

    1999-04-01

    The use of Particle Induced X-ray Emission (PIXE), Rutherford Backscattering Spectrometry (RBS) and Scanning Transmission Ion Microscopy (STIM) to provide quantitative elemental analysis of single cells is an area which has high potential, particularly when the trace elements such as Ca, Fe, Zn and Cu can be monitored. We describe the methodology of sample preparation for two cell types, the procedures of cell imaging using STIM, and the quantitative elemental analysis of single cells using RBS and PIXE. Recent work on single cells at the Nuclear Microscopy Research Centre,National University of Singapore has centred around two research areas: (a) Apoptosis (programmed cell death), which has been recently implicated in a wide range of pathological conditions such as cancer, Parkinson's disease etc, and (b) Malaria (infection of red blood cells by the malaria parasite). Firstly we present results on the elemental analysis of human Chang liver cells (ATTCC CCL 13) where vanadium ions were used to trigger apoptosis, and demonstrate that nuclear microscopy has the capability of monitoring vanadium loading within individual cells. Secondly we present the results of elemental changes taking place in individual mouse red blood cells which have been infected with the malaria parasite and treated with the anti-malaria drug Qinghaosu (QHS).

  1. Nuclear fission and the transuranium elements

    SciTech Connect

    Seaborg, G.T.

    1989-02-01

    Many of the transuranium elements are produced and isolated in large quantities through the use of neutrons furnished by nuclear fission reactions: plutonium (atomic number 94) in ton quantities; neptunium (93), americium (95), and curium (96) in kilogram quantities; berkelium (97) in 100 milligram quantities; californium (98) in gram quantities; and einsteinium (99) in milligram quantities. Transuranium isotopes have found many practical applications---as nuclear fuel for the large-scale generation of electricity, as compact, long-lived power sources for use in space exploration, as means for diagnosis and treatment in the medical area, and as tools in numerous industrial processes. Of particular interest is the unusual chemistry and impact of these heaviest elements on the periodic table. This account will feature these aspects. 9 refs., 5 figs.

  2. Nuclear microscopy of sperm cell elemental structure

    NASA Astrophysics Data System (ADS)

    Bench, Graham S.; Balhorn, Rod; Friz, Alexander M.

    1995-05-01

    Theories suggest there is a link between protamine concentrations in individual sperm and male fertility. Previously, biochemical analyses have used pooled samples containing millions of sperm to determine protamine concentrations. These methods have not been able to determine what percentage of morphologically normal sperm are biochemically defective and potentially infertile. Nuclear microscopy has been utilized to measure elemental profiles at the single sperm level. By measuring the amount of phosphorus and sulfur, the total DNA and protamine content in individual sperm from fertile bull and mouse semen have been determined. These values agree with results obtained from other biochemical analyses. Nuclear microscopy shows promise for measuring elemental profiles in the chromatin of individual sperm. The technique may be able to resolve theories regarding the importance of protamines to male fertility and identify biochemical defects responsible for certain types of male infertility.

  3. Nuclear microscopy of sperm cell elemental structure

    SciTech Connect

    Bench, G.S.; Balhorn, R.; Friz, A.M.; Freeman, S.P.H.T.

    1994-09-28

    Theories suggest there is a link between protamine concentrations in individual sperm and male fertility. Previously, biochemical analyses have used pooled samples containing millions of sperm to determine protamine concentrations. These methods have not been able to determine what percentage of morphologically normal sperm are biochemically defective and potentially infertile. Nuclear microscopy has been utilized to measure elemental profiles at the single sperm level. By measuring the amount of phosphorus and sulfur, the total DNA and protamine content in individual sperm from fertile bull and mouse semen have been determined. These values agree with results obtained from other biochemical analyses. Nuclear microscopy shows promise for measuring elemental profiles in the chromatin of individual sperm. The technique may be able to resolve theories regarding the importance of protamines to male fertility and identify biochemical defects responsible for certain types of male infertility.

  4. NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR

    DOEpatents

    Rasor, N.S.; Hirsch, R.L.

    1963-12-01

    The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

  5. 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.

  6. 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.

  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. Nuclear Reactions Used For Superheavy Element Research

    SciTech Connect

    Stoyer, Mark A.

    2008-04-17

    Some of the most fascinating questions about the limits of nuclear stability are confronted in the heaviest nuclei. How many more new elements can be synthesized? What are the nuclear and chemical properties of these exotic nuclei? Does the 'Island of Stability' exist and can we ever explore the isotopes inhabiting that nuclear region? This paper will focus on the current experimental research on the synthesis and characterization of superheavy nuclei with Z>112 from the Dubna/Livermore collaboration. Reactions using {sup 48}Ca projectiles from the U400 cyclotron and actinide targets ({sup 233,238}U, {sup 237}Np, {sup 242,244}Pu, {sup 243}Am, {sup 245,248}Cm, {sup 249}Cf) have been investigated using the Dubna Gas Filled Recoil Separator in Dubna over the last 8 years. In addition, several experiments have been performed to investigate the chemical properties of some of the observed longer-lived isotopes produced in these reactions. Some comments will be made on nuclear reactions used for the production of the heaviest elements. A summary of the current status of the upper end of the chart of nuclides will be presented.

  9. Nuclear reactions used for superheavy element research

    SciTech Connect

    Stoyer, M A

    2008-02-26

    Some of the most fascinating questions about the limits of nuclear stability are confronted in the heaviest nuclei. How many more new elements can be synthesized? What are the nuclear and chemical properties of these exotic nuclei? Does the 'Island of Stability' exist and can we ever explore the isotopes inhabiting that nuclear region? This paper will focus on the current experimental research on the synthesis and characterization of superheavy nuclei with Z > 112 from the Dubna/Livermore collaboration. Reactions using 48Ca projectiles from the U400 cyclotron and actinide targets ({sup 233,238}U, {sup 237}Np, {sup 242,244}Pu, {sup 243}Am, {sup 245,248}Cm, {sup 249}Cf) have been investigated using the Dubna Gas Filled Recoil Separator in Dubna over the last 8 years. In addition, several experiments have been performed to investigate the chemical properties of some of the observed longer-lived isotopes produced in these reactions. Some comments will be made on nuclear reactions used for the production of the heaviest elements. A summary of the current status of the upper end of the chart of nuclides will be presented.

  10. Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2008-01-01

    To support the eventual development of a nuclear thermal rocket engine, a state-of-the-art experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The test device simulates the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner as to accurately reproduce the temperatures and heat fluxes normally expected to occur as a result of nuclear fission while at the same time being exposed to flowing hydrogen. This project is referred to as the Nuclear Thermal Rocket Element Environment Simulator or NTREES. The NTREES device is located at the Marshall Space flight Center in a laboratory which has been modified to accommodate the high powers required to heat the test articles to the required temperatures and to handle the gaseous hydrogen flow required for the tests. Other modifications to the laboratory include the installation of a nitrogen gas supply system and a cooling water supply system. During the design and construction of the facility, every effort was made to comply with all pertinent regulations to provide assurance that the facility could be operated in a safe and efficient manner. The NTREES system can currently supply up to 50 kW of inductive heating to the fuel test articles, although the facility has been sized to eventually allow test article heating levels of up to several megawatts.

  11. Nuclear fuel elements made from nanophase materials

    DOEpatents

    Heubeck, Norman B.

    1998-01-01

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000.degree. F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics.

  12. Nuclear fuel elements made from nanophase materials

    DOEpatents

    Heubeck, N.B.

    1998-09-08

    A nuclear reactor core fuel element is composed of nanophase high temperature materials. An array of the fuel elements in rod form are joined in an open geometry fuel cell that preferably also uses such nanophase materials for the cell structures. The particular high temperature nanophase fuel element material must have the appropriate mechanical characteristics to avoid strain related failure even at high temperatures, in the order of about 3000 F. Preferably, the reactor type is a pressurized or boiling water reactor and the nanophase material is a high temperature ceramic or ceramic composite. Nanophase metals, or nanophase metals with nanophase ceramics in a composite mixture, also have desirable characteristics, although their temperature capability is not as great as with all-ceramic nanophase material. Combinations of conventional or nanophase metals and conventional or nanophase ceramics can be employed as long as there is at least one nanophase material in the composite. The nuclear reactor so constructed has a number of high strength fuel particles, a nanophase structural material for supporting a fuel rod at high temperature, a configuration to allow passive cooling in the event of a primary cooling system failure, an ability to retain a coolable geometry even at high temperatures, an ability to resist generation of hydrogen gas, and a configuration having good nuclear, corrosion, and mechanical characteristics. 5 figs.

  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. 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,…

  16. 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

  17. 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.

  18. 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.

  19. Element speciation during nuclear glass alteration

    NASA Astrophysics Data System (ADS)

    Galoisy, L.; Calas, G.; Bergeron, B.; Jollivet, P.; Pelegrin, E.

    2011-12-01

    Assessing the long-term behavior of nuclear glasses implies the prediction of their long-term performance. An important controlling parameter is their evolution during interaction with water under conditions simulating geological repositories. After briefly recalling the major characteristics of the local and medium-range structure of borosilicate glasses of nuclear interest, we will present some structural features of this evolution. Specific structural tools used to determine the local structure of glass surfaces include synchrotron-radiation x-ray absorption spectroscopy with total electron yield detection. The evolution of the structure of glass surface has been determined at the Zr-, Fe-, Si- and Al-K edges and U-LIII edge. During alteration in near- or under-saturated conditions, some elements such as Fe change coordination, as other elements such as Zr only suffer structural modifications in under-saturated conditions. Uranium exhibits a modification of its speciation from an hexa-coordinated U(VI) in the borosilicate glass to an uranyl group in the gel. These structural modifications may explain the chemical dependence of the initial alteration rate and the transition to the residual regime. They also illustrate the molecular-scale origin of the processes at the origin of the glass-to-gel transformation. Eventually, they explain the provisional trapping of U by the alteration gel: the uranium retention factors in the gel depend on the alteration conditions, and thus on the nature of the resulting gel and on the trapping conditions.

  20. Tracing nuclear elements released by Fukushima Nuclear Power Plant accident

    NASA Astrophysics Data System (ADS)

    Tsujimura, M.; Onda, Y.; Abe, Y.; Hada, M.; Pun, I.

    2011-12-01

    Radioactive contamination has been detected in Fukushima and the neighboring regions due to the nuclear accident at Fukushima Daiichi Nuclear Power Plant (NPP) following the earthquake and tsunami occurred on 11th March 2011. The small experimental catchments have been established in Yamakiya district, Kawamata Town, Fukushima Prefecture, located approximately 35 km west from the Fukushima NPP. The tritium (3H) concentration and stable isotopic compositions of deuterium and oxygen-18 have been determined on the water samples of precipitation, soil water at the depths of 10 to 30 cm, groundwater at the depths of 5 m to 50 m, spring water and stream water taken at the watersheds in the recharge and discharge zones from the view point of the groundwater flow system. The tritium concentration of the rain water fell just a few days after the earthquake showed a value of approximately 17 Tritium Unit (T.U.), whereas the average concentration of the tritium in the precipitation was less than 5 T.U. before the Fukushima accident. The spring water in the recharge zone showed a relatively high tritium concentration of approximately 12 T.U., whereas that of the discharge zone showed less than 5 T.U. Thus, the artificial tritium was apparently injected in the groundwater flow system due to the Fukushima NPP accident, whereas that has not reached at the discharge zone yet. The monitoring of the nuclear elements is now on going from the view points of the hydrological cycles and the drinking water security.

  1. 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...

  2. Drive of nuclear reactor's control element

    SciTech Connect

    Anikin, A.A.; But, V.G.; Nikolaev, V.P.; Silvanovich, A.A.

    1980-12-09

    According to the invention, the drive of a nuclear reactor's control element comprises an electromotor having a stator and a rotor composed lengthwise of two parts whose total length is equal to that of the active part of the stator. One part of the rotor is a solid cylinder-shaped member. The other part of the rotor comprises at least three double-arm rocking levers, the pivot axes of which are parallel to the axis of a drive screw. One arm of each of said levers is a rotor pole. The other arm of each of said levers carries a roller, the axis of rotation of which is parallel to the axis of the drive screw. Said rollers make up a detachable roller nut which interacts with the drive screw under the action of an electromagnetic field.

  3. Nuclear microscopy of sperm cell elemental structure

    SciTech Connect

    Bench, G.S.

    1994-12-31

    Theories have suggested that there is a link between protamine concentrations in individual sperm and sperm fertility. At present, biochemical analyses have only been performed on bulk populations and existing methods have not been able to determine what percentage of morphologically normal sperm are biochemically defective and potentially infertile. As part of an investigation into male sperm fertility, nuclear microscopy has been utilized to measure elemental profiles at the single sperm level. By measuring the ratio of Phosphorus to Sulfur the authors have been able to determine the amount of protamine 1 and protamine 2 in individual cells from bulk fertile samples of bull and mouse sperm. Preliminary results show that, for each species, the relative amounts of protamine 1 and protamine 2 in morphologically normal sperm agree well with expected values.

  4. Fuel element concept for long life high power nuclear reactors

    NASA Technical Reports Server (NTRS)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  5. 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.

  6. 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.

  7. 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...

  8. 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...

  9. 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...

  10. 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...

  11. 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...

  12. 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…

  13. 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.

  14. 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…

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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…

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Process to separate transuranic elements from nuclear waste

    DOEpatents

    Johnson, Terry R.; Ackerman, John P.; Tomczuk, Zygmunt; Fischer, Donald F.

    1989-01-01

    A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

  8. 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.

  9. 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)

  10. 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.

  11. 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...

  12. 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.

  13. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2014-01-01

    Over the past year the Nuclear Thermal Rocket Element Environmental Simulator (NTREES) has been undergoing a significant upgrade beyond its initial configuration. The NTREES facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The first phase of the upgrade activities which was completed in 2012 in part consisted of an extensive modification to the hydrogen system to permit computer controlled operations outside the building through the use of pneumatically operated variable position valves. This setup also allows the hydrogen flow rate to be increased to over 200 g/sec and reduced the operation complexity of the system. The second stage of modifications to NTREES which has just been completed expands the capabilities of the facility significantly. In particular, the previous 50 kW induction power supply has been replaced with a 1.2 MW unit which should allow more prototypical fuel element temperatures to be reached. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during. This new setup required that the NTREES vessel be raised onto a platform along with most of its associated gas and vent lines. In this arrangement, the induction heater and water systems are now located underneath the platform. In this new configuration, the 1.2 MW NTREES induction heater will be capable of testing fuel elements and fuel materials in flowing hydrogen at pressures up to 1000 psi at temperatures up to and beyond 3000 K and at near-prototypic reactor channel power densities. NTREES is also capable of testing potential fuel elements with a variety of propellants, including hydrogen with additives to inhibit

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    NASA Technical Reports Server (NTRS)

    Emrich, William J. Jr.; Moran, Robert P.; Pearson, J. Boise

    2012-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities

  19. Nuclear structure notes on element 115 decay chains

    SciTech Connect

    Rudolph, D. Sarmiento, L. G.; Forsberg, U.

    2015-10-15

    Hitherto collected data on more than hundred α-decay chains stemming from element 115 are combined to probe some aspects of the underlying nuclear structure of the heaviest atomic nuclei yet created in the laboratory.

  20. 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.

  1. Chem I Supplement: Nuclear Synthesis and Identification of New Elements.

    ERIC Educational Resources Information Center

    Seaborg, Glenn T.

    1985-01-01

    As background material for a paper on the transuranium elements (SE 537 837), this article reviews: (1) several descriptive terms; (2) nuclear reactions; (3) radioactive decay modes; (4) chemical background; and (5) experimental methods used in this field of research and more broadly in nuclear chemistry. (Author/JN)

  2. 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.

  3. 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

  4. Neutrinoless Double Beta Nuclear Matrix Elements Around Mass 80 in the Nuclear Shell Model

    NASA Astrophysics Data System (ADS)

    Yoshinaga, Naotaka; Higashiyama, Koji; Taguchi, Daisuke; Teruya, Eri

    The observation of the neutrinoless double-beta decay can determine whether the neutrino is a Majorana particle or not. In its theoretical nuclear side it is particularly important to estimate three types of nuclear matrix elements, namely, Fermi (F), Gamow-Teller (GT), and tensor (T) types matrix elements. The shell model calculations and also the pair-truncated shell model calculations are carried out to check the model dependence on nuclear matrix elements. In this work the neutrinoless double-beta decay for mass A = 82 nuclei is studied. It is found that the matrix elements are quite sensitive to the ground state wavefunctions.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. Process to separate transuranic elements from nuclear waste

    DOEpatents

    Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

    1989-03-21

    A process is described for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

  11. Process to separate transuranic elements from nuclear waste

    DOEpatents

    Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

    1988-07-12

    A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

  12. 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.

  13. Civil defense: nuclear debate's new element

    SciTech Connect

    Sweet, W.

    1982-06-04

    President Reagan's plans to build up US military forces include a seven-year $4.2 billion civil-defense program that will emphasize the removal of residents from urban centers and will match the Soviet capability. In the Soviet Union, cities of 100,000 have shelters for 10 to 20% of the population, but they lack the US transportation system, low-density suburbs, mild climate, and other factors. The House Armed Services Committee approved raising the current $133 million civil-defense budget to $252 million, but the Senate's May 14th vote limited the increase to $144 million. The civil-defense debate offers peace and anti-nuclear activists on opportunity to organize and coordinate their efforts. Peace activists were to demonstrate against the administration's plans during a special June 7-9 United Nations session because they feel the public will now be able to understand the implications of relocation in government planning for nuclear war. 17 references, 2 figures, 31 tables. (DCK)

  14. 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.

  15. 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).

  16. 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

  17. 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.

  18. Means for supporting fuel elements in a nuclear reactor

    DOEpatents

    Andrews, Harry N.; Keller, Herbert W.

    1980-01-01

    A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

  19. 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.

  20. 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.

  1. 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…

  2. 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).

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. Superheavy Element Nuclear Chemistry at RIKEN

    SciTech Connect

    Haba, Hiromitsu; Kaji, Daiya; Kasamatsu, Yoshitaka; Kudou, Yuki; Morimoto, Kouji; Morita, Kosuke; Ozeki, Kazutaka; Yoneda, Akira; Kikunaga, Hidetoshi; Komori, Yukiko; Ooe, Kazuhiro; Shinohara, Atsushi; Yoshimura, Takashi; Sato, Nozomi; Toyoshima, Atsushi; Yokoyama, Akihiko

    2010-05-12

    A gas-jet transport system has been coupled to the RIKEN gas-filled recoil ion separator GARIS to startup superheavy element (SHE) chemistry at RIKEN. The performance of the system was appraised using an isotope of element 104, {sup 261}Rf, produced in the {sup 248}Cm({sup 18}O,5n){sup 261}Rf reaction. Alpha-particles of {sup 261}Rf separated with GARIS and extracted to a chemistry laboratory were successfully identified with a rotating wheel apparatus for alpha spectrometry. The setting parameters such as the magnetic field of the separator and the gas-jet conditions were optimized. The present results suggest that the GARIS/gas-jet system is a promising approach for exploring new frontiers in SHE chemistry: (i) the background radioactivities of unwanted reaction products are strongly suppressed, (ii) the intense beam is absent in the gas-jet chamber and hence high gas-jet efficiency is achieved, and (iii) the beam-free condition also allows for investigations of new chemical systems.

  8. 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.

  9. Direct mapping of nuclear shell effects in the heaviest elements.

    PubMed

    Minaya Ramirez, E; Ackermann, D; Blaum, K; Block, M; Droese, C; Düllmann, Ch E; Dworschak, M; Eibach, M; Eliseev, S; Haettner, E; Herfurth, F; Heßberger, F P; Hofmann, S; Ketelaer, J; Marx, G; Mazzocco, M; Nesterenko, D; Novikov, Yu N; Plaß, W R; Rodríguez, D; Scheidenberger, C; Schweikhard, L; Thirolf, P G; Weber, C

    2012-09-07

    Quantum-mechanical shell effects are expected to strongly enhance nuclear binding on an "island of stability" of superheavy elements. The predicted center at proton number Z = 114, 120, or 126 and neutron number N = 184 has been substantiated by the recent synthesis of new elements up to Z = 118. However, the location of the center and the extension of the island of stability remain vague. High-precision mass spectrometry allows the direct measurement of nuclear binding energies and thus the determination of the strength of shell effects. Here, we present such measurements for nobelium and lawrencium isotopes, which also pin down the deformed shell gap at N = 152.

  10. 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.

  11. 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...

  12. FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS

    DOEpatents

    Flint, O.

    1961-01-10

    Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

  13. 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.

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Layer, E.H. Jr.; Peet, C.S.

    1962-01-23

    A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

  19. Multi-Elemental Nuclear Analysis of soil reference material

    NASA Astrophysics Data System (ADS)

    Metairon, S.; Zamboni, C. B.; Medeiros, I. M. M. Amaral; Menezes, M. À. B. C.

    2011-08-01

    The elements concentration in the soil reference material (IAEA/SOIL-7) was obtained using the parametric Neutron Activation Analysis technique in the IEA-R1 nuclear reactor at IPEN (CNEN-SP). The results obtained were in good agreement with the respective nominal values from this reference material suggesting the viability of using this parametric procedure for environmental investigations.

  20. METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Hauth, J.J.; Anicetti, R.J.

    1962-12-01

    A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

  1. 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.

  2. 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.

  3. 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.

  4. 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...

  5. 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...

  6. 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.

  7. Neutrinoless double-β decay and nuclear transition matrix elements

    SciTech Connect

    Rath, P. K.

    2015-10-28

    Within mechanisms involving the light Majorana neutrinos, squark-neutrino, Majorons, sterile neutrinos and heavy Majorana neutrino, nuclear transition matrix elements for the neutrinoless (β{sup −}β{sup −}){sub 0ν} decay of {sup 96}Zr, {sup 100}Mo, {sup 128,130}Te and {sup 150}Nd nuclei are calculated by employing the PHFB approach. Effects due to finite size of nucleons, higher order currents, short range correlations, and deformations of parent as well as daughter nuclei on the calculated matrix elements are estimated. Uncertainties in nuclear transition matrix elements within long-ranged mechanisms but for double Majoron accompanied (β{sup −}β{sup −}ϕϕ){sub 0ν} decay modes are 9%–15%. In the case of short ranged heavy Majorona neutrino exchange mechanism, the maximum uncertainty is about 35%. The maximum systematic error within the mechanism involving the exchange of light Majorana neutrino is about 46%.

  8. Methods for making a porous nuclear fuel element

    DOEpatents

    Youchison, Dennis L; Williams, Brian E; Benander, Robert E

    2014-12-30

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  9. 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.

  10. 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.

  11. 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.

  12. 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,...

  13. 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 ...

  14. Initial Operation of the Nuclear Thermal Rocket Element Environmental Simulator

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.; Pearson, J. Boise; Schoenfeld, Michael P.

    2015-01-01

    The Nuclear Thermal Rocket Element Environmental Simulator (NTREES) facility is designed to perform realistic non-nuclear testing of nuclear thermal rocket (NTR) fuel elements and fuel materials. Although the NTREES facility cannot mimic the neutron and gamma environment of an operating NTR, it can simulate the thermal hydraulic environment within an NTR fuel element to provide critical information on material performance and compatibility. The NTREES facility has recently been upgraded such that the power capabilities of the facility have been increased significantly. At its present 1.2 MW power level, more prototypical fuel element temperatures nay now be reached. The new 1.2 MW induction heater consists of three physical units consisting of a transformer, rectifier, and inverter. This multiunit arrangement facilitated increasing the flexibility of the induction heater by more easily allowing variable frequency operation. Frequency ranges between 20 and 60 kHz can accommodated in the new induction heater allowing more representative power distributions to be generated within the test elements. The water cooling system was also upgraded to so as to be capable of removing 100% of the heat generated during testing In this new higher power configuration, NTREES will be capable of testing fuel elements and fuel materials at near-prototypic power densities. As checkout testing progressed and as higher power levels were achieved, several design deficiencies were discovered and fixed. Most of these design deficiencies were related to stray RF energy causing various components to encounter unexpected heating. Copper shielding around these components largely eliminated these problems. Other problems encountered involved unexpected movement in the coil due to electromagnetic forces and electrical arcing between the coil and a dummy test article. The coil movement and arcing which were encountered during the checkout testing effectively destroyed the induction coil in use at

  15. 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.

  16. Nuclear fuel elements and method of making same

    DOEpatents

    Schweitzer, Donald G.

    1992-01-01

    A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

  17. 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.

  18. 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.

  19. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    DOEpatents

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  20. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Upgrade Activities

    NASA Technical Reports Server (NTRS)

    Emrich, William

    2013-01-01

    A key technology element in Nuclear Thermal Propulsion is the development of fuel materials and components which can withstand extremely high temperatures while being exposed to flowing hydrogen. NTREES provides a cost effective method for rapidly screening of candidate fuel components with regard to their viability for use in NTR systems. The NTREES is designed to mimic the conditions (minus the radiation) to which nuclear rocket fuel elements and other components would be subjected to during reactor operation. The NTREES consists of a water cooled ASME code stamped pressure vessel and its associated control hardware and instrumentation coupled with inductive heaters to simulate the heat provided by the fission process. The NTREES has been designed to safely allow hydrogen gas to be injected into internal flow passages of an inductively heated test article mounted in the chamber.

  1. Modeling and Simulation of a Nuclear Fuel Element Test Section

    NASA Technical Reports Server (NTRS)

    Moran, Robert P.; Emrich, William

    2011-01-01

    "The Nuclear Thermal Rocket Element Environmental Simulator" test section closely simulates the internal operating conditions of a thermal nuclear rocket. The purpose of testing is to determine the ideal fuel rod characteristics for optimum thermal heat transfer to their hydrogen cooling/working fluid while still maintaining fuel rod structural integrity. Working fluid exhaust temperatures of up to 5,000 degrees Fahrenheit can be encountered. The exhaust gas is rendered inert and massively reduced in temperature for analysis using a combination of water cooling channels and cool N2 gas injectors in the H2-N2 mixer portion of the test section. An extensive thermal fluid analysis was performed in support of the engineering design of the H2-N2 mixer in order to determine the maximum "mass flow rate"-"operating temperature" curve of the fuel elements hydrogen exhaust gas based on the test facilities available cooling N2 mass flow rate as the limiting factor.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    NASA Technical Reports Server (NTRS)

    Bradley, D. E.; Mireles, O. R.; Hickman, R. R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse and relatively high thrust to achieve mission goals in reasonable time frames.1,2 Conventional storable propellants produce average specific impulse. Nuclear thermal rockets capable of producing high specific impulse are proposed. Nuclear thermal rockets employ heat produced by fission reaction to heat and therefore accelerate hydrogen, which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000 K), and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high-temperature hydrogen exposure on fuel elements are limited.3 The primary concern is the mechanical failure of fuel elements that employ high-melting-point metals, ceramics, or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. The purpose of the testing is to obtain data to assess the properties of the non-nuclear support materials, as-fabricated, and determine their ability to survive and maintain thermal performance in a prototypical NTR reactor environment of exposure to hydrogen at very high temperatures. The fission process of the planned fissile material and the resulting heating performance is well known and does not therefore require that active fissile material be integrated in this testing. A small-scale test bed designed to heat fuel element samples via non-contact radio frequency heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  8. 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.

  9. Neutrinoless double beta nuclear matrix elements around mass 80 in the nuclear shell-model

    NASA Astrophysics Data System (ADS)

    Yoshinaga, N.; Higashiyama, K.; Taguchi, D.; Teruya, E.

    2015-05-01

    The observation of the neutrinoless double-beta decay can determine whether the neutrino is a Majorana particle or not. For theoretical nuclear physics it is particularly important to estimate three types of matrix elements, namely Fermi (F), Gamow-Teller (GT), and tensor (T) matrix elements. In this paper, we carry out shell-model calculations and also pair-truncated shell-model calculations to check the model dependence in the case of mass A=82 nuclei.

  10. ANSI/ANS-8.15-1981(R87): Nuclear criticality control of special actinide elements

    SciTech Connect

    Brewer, R.W.; Pruvost, N.L.; Rombough, C.T.

    1996-12-31

    The American National Standard, {open_quotes}Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactotors{close_quotes} American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1-1983(R88) provides guidance for the nuclides {sup 233}U, {sup 235}U, and {sup 239}Pu. These three nuclides are of primary interest in out-of-reactor criticality safety since they are the most commonly encountered in the vast majority of operations. However, some operations can involve nuclides other than {sup 233}U, {sup 235}U, and {sup 239}Pu in sufficient quantities that their effect on criticality safety could be of concern. ANSI/ANS-8.15-1981(R87) {open_quotes}Nuclear Criticality Control of Special Actinide Elements,{close_quotes} provides guidance for 15 such nuclides. The standard was approved for use on November 9, 1981. When it received its first 5-yr review, no changes were made, and it was reaffirmed effective October 30, 1987. The standard was again reviewed and reaffirmed without changes in December 1995. The next 5-yr review of the standard is due in December 2000. The affected nuclides are {sup 237}Np, {sup 238}Pu, {sup 240}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm, {sup 239}Pu, {sup 241}Pu, {sup 242m}Am, {sup 243}Cm, {sup 245}Cm, {sup 247}Cm, {sup 249}Cf, and {sup 251}Cf.

  11. Low Cost Nuclear Thermal Rocket Cermet Fuel Element Environment Testing

    NASA Technical Reports Server (NTRS)

    Bradley, David E.; Mireles, Omar R.; Hickman, Robert R.

    2011-01-01

    Deep space missions with large payloads require high specific impulse (Isp) and relatively high thrust in order to achieve mission goals in reasonable time frames. Conventional, storable propellants produce average Isp. Nuclear thermal rockets (NTR) capable of high Isp thrust have been proposed. NTR employs heat produced by fission reaction to heat and therefore accelerate hydrogen which is then forced through a rocket nozzle providing thrust. Fuel element temperatures are very high (up to 3000K) and hydrogen is highly reactive with most materials at high temperatures. Data covering the effects of high temperature hydrogen exposure on fuel elements is limited. The primary concern is the mechanical failure of fuel elements which employ high-melting-point metals, ceramics or a combination (cermet) as a structural matrix into which the nuclear fuel is distributed. It is not necessary to include fissile material in test samples intended to explore high temperature hydrogen exposure of the structural support matrices. A small-scale test bed designed to heat fuel element samples via non-contact RF heating and expose samples to hydrogen is being developed to assist in optimal material and manufacturing process selection without employing fissile material. This paper details the test bed design and results of testing conducted to date.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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...

  18. 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...

  19. 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

  20. Methods for manufacturing porous nuclear fuel elements for high-temperature gas-cooled nuclear reactors

    DOEpatents

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2010-02-23

    Methods for manufacturing porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's). Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, a thin coating of nuclear fuel may be deposited inside of a highly porous skeletal structure made, for example, of reticulated vitreous carbon foam.

  1. Porous nuclear fuel element with internal skeleton for high-temperature gas-cooled nuclear reactors

    DOEpatents

    Youchison, Dennis L.; Williams, Brian E.; Benander, Robert E.

    2013-09-03

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  2. Porous nuclear fuel element for high-temperature gas-cooled nuclear reactors

    DOEpatents

    Youchison, Dennis L [Albuquerque, NM; Williams, Brian E [Pacoima, CA; Benander, Robert E [Pacoima, CA

    2011-03-01

    Porous nuclear fuel elements for use in advanced high temperature gas-cooled nuclear reactors (HTGR's), and to processes for fabricating them. Advanced uranium bi-carbide, uranium tri-carbide and uranium carbonitride nuclear fuels can be used. These fuels have high melting temperatures, high thermal conductivity, and high resistance to erosion by hot hydrogen gas. Tri-carbide fuels, such as (U,Zr,Nb)C, can be fabricated using chemical vapor infiltration (CVI) to simultaneously deposit each of the three separate carbides, e.g., UC, ZrC, and NbC in a single CVI step. By using CVI, the nuclear fuel may be deposited inside of a highly porous skeletal structure made of, for example, reticulated vitreous carbon foam.

  3. 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

  4. 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.

  5. Nuclear reactor fuel element with vanadium getter on cladding

    DOEpatents

    Johnson, Carl E.; Carroll, Kenneth G.

    1977-01-01

    A nuclear reactor fuel element is described which has an outer cladding, a central core of fissionable or mixed fissionable and fertile fuel material and a layer of vanadium as an oxygen getter on the inner surface of the cladding. The vanadium reacts with oxygen released by the fissionable material during irradiation of the core to prevent the oxygen from reacting with and corroding the cladding. Also described is a method for coating the inner surface of small diameter tubes of cladding with a layer of vanadium.

  6. 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.

  7. Nuclear reactor fuel element having improved heat transfer

    DOEpatents

    Garnier, J.E.; Begej, S.; Williford, R.E.; Christensen, J.A.

    1982-03-03

    A nuclear reactor fuel element having improved heat transfer between fuel material and cladding is described. The element consists of an outer cladding tube divided into an upper fuel section containing a central core of fissionable or mixed fissionable and fertile fuel material, slightly smaller in diameter than the inner surface of the cladding tube and a small lower accumulator section, the cladding tube being which is filled with a low molecular weight gas to transfer heat from fuel material to cladding during irradiation. A plurality of essentially vertical grooves in the fuel section extend downward and communicate with the accumulator section. The radial depth of the grooves is sufficient to provide a thermal gradient between the hot fuel surface and the relatively cooler cladding surface to allow thermal segregation to take place between the low molecular weight heat transfer gas and high molecular weight fission product gases produced by the fuel material during irradiation.

  8. 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.

  9. 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.

  10. 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.

  11. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

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

  12. 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....

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

    Code of Federal Regulations, 2010 CFR

    2010-01-01

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

  14. 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....

  15. 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...

  16. 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...

  17. 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...

  18. 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...

  19. 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...

  20. 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.

  1. 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.

  2. Safety Education--An Essential Element of Technical Training

    ERIC Educational Resources Information Center

    Glazener, Everett R.; Comstock, Thomas W.

    1978-01-01

    After tracing the background of the safety movement, provisions of recent safety legislation, and the ecological and environmental impact of industrial processes, the author stresses the need for safety education in technical education programs to prepare future employees for industrial safety requirements. (MF)

  3. 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).

  4. 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.

  5. American National Standard ANSI/ANS-8.15-1983: Nuclear criticality control of special actinide elements

    SciTech Connect

    Brewer, R.W.; Pruvost, N.L.; Rombough, C.T.

    1996-12-31

    The American National Standard, `Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors` ANSI/ANS-8.1- 1983 provides guidance for the nuclides [sup 233]U, [sup 235]U, and [sup 239]Pu These three nuclides are of primary interest in out-of-reactor criticality safety since they are the most commonly encountered in the vast majority of operations. However, some operations can involve nuclides other than `U, `U, and `Pu in sufficient quantities that their effect on criticality safety could be of concern. The American National Standard, `Nuclear Criticality Control of Special Actinide Elements` ANSI/ANS-8.`15-1983 (Ref 2), provides guidance for fifteen such nuclides.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  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

    1986-07-01

    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.

  11. 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.

  12. 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,...

  13. 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...

  14. 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,...

  15. 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,...

  16. 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,...

  17. 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,...

  18. 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.

  19. 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.

  20. Thermo-Elastic Finite Element Analyses of Annular Nuclear Fuels

    NASA Astrophysics Data System (ADS)

    Kwon, Y. D.; Kwon, S. B.; Rho, K. T.; Kim, M. S.; Song, H. J.

    In this study, we tried to examine the pros and cons of the annular type of fuel concerning mainly with the temperatures and stresses of pellet and cladding. The inner and outer gaps between pellet and cladding may play an important role on the temperature distribution and stress distribution of fuel system. Thus, we tested several inner and outer gap cases, and we evaluated the effect of gaps on fuel systems. We conducted thermo-elastic-plastic-creep analyses using an in-house thermo-elastic-plastic-creep finite element program that adopted the 'effective-stress-function' algorithm. Most analyses were conducted until the gaps disappeared; however, certain analyses lasted for 1582 days, after which the fuels were replaced. Further study on the optimal gaps sizes for annular nuclear fuel systems is still required.

  1. TFIIIC Bound DNA Elements in Nuclear Organization and Insulation

    PubMed Central

    Kirkland, Jacob G.; Raab, Jesse R.

    2012-01-01

    tRNA genes (tDNAs) have been known to have barrier insulator function in budding yeast, Saccharomyces cerevisiae, for over a decade. tDNAs also play a role in genome organization by clustering at sites in the nucleus and both of these functions are dependent on the transcription factor TFIIIC. More recently TFIIIC bound sites devoid of pol III, termed Extra-TFIIIC sites (ETC) have been identified in budding yeast and these sites also function as insulators and affect genome organization. Subsequent studies in Schizosaccharomyces pombe showed that TFIIIC bound sites were insulators and also functioned as Chromosome Organization Clamps (COC); tethering the sites to the nuclear periphery. Very recently studies have moved to mammalian systems where pol III genes and their associated factors have been investigated in both mouse and human cells. Short Interspersed Nuclear Elements (SINEs) that bind TFIIIC, function as insulator elements and tDNAs can also function as both enhancer -blocking and barrier insulators in these organisms. It was also recently shown that tDNAs cluster with other tDNAs and with ETCs but not with pol II transcribed genes. Intriguingly, TFIIIC is often found near pol II transcription start sites and it remains unclear what the consequences of TFIIIC based genomic organization are and what influence pol III factors have on pol II transcribed genes and vise versa. In this review we provide a comprehensive overview of the known data on pol III factors in insulation and genome organization and identify the many open questions that require further investigation. \\ PMID:23000638

  2. 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.

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

    NASA Astrophysics Data System (ADS)

    Baranko, Benjamin T.

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

  4. 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.

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

    SciTech Connect

    1983-02-01

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

  6. 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

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

    SciTech Connect

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

    2006-07-01

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

  8. 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.

  9. 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...

  10. 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.

  11. 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.

  12. Conditions for the successful integration of Human and Organizational Factors (HOF) in the nuclear safety analysis.

    PubMed

    Tosello, Michèle; Lévêque, Françoise; Dutillieu, Stéphanie; Hernandez, Guillaume; Vautier, Jean-François

    2012-01-01

    This communication presents some elements which come from the experience feedback at CEA about the conditions for the successful integration of HOF in the nuclear safety analysis. To point out some of these conditions, one of the concepts proposed by Edgar Morin to describe the functioning of "complex" systems: the dialogical principle has been used. The idea is to look for some dialogical pairs. The elements of this kind of pair are both complementary and antagonist to one another. Three dialogical pairs are presented in this communication. The first two pairs are related to the organization of the HOF network and the last one is related to the methods which are used to analyse the working situations. The three pairs are: specialist - non-specialist actors of the network, centralized - distributed human resources in the network and microscopic - macroscopic levels of HOF methods to analyse the working situations. To continuously improve these three dialogical pairs, it is important to keep the differences which exist between the two elements of a pair and to find and maintain a balance between the two elements of the pairs.

  13. Nuclear criticality safety modeling of an LEU deposit

    SciTech Connect

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

    1996-11-01

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

  14. Nuclear criticality safety program for environmental restoration projects

    SciTech Connect

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

    1994-05-01

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

  15. Nuclear physics and heavy element research at LLNL

    SciTech Connect

    Stoyer, M A; Ahle, L E; Becker, J A; Bernstein, L A; Bleuel, D L; Burke, J T; Dashdorj, D; Henderson, R A; Hurst, A M; Kenneally, J M; Lesher, S R; Moody, K J; Nelson, S L; Norman, E B; Pedretti, M; Scielzo, N D; Shaughnessy, D A; Sheets, S A; Stoeffl, W; Stoyer, N J; Wiedeking, M; Wilk, P A; Wu, C Y

    2009-05-11

    This paper highlights some of the current basic nuclear physics research at Lawrence Livermore National Laboratory (LLNL). The work at LLNL concentrates on investigating nuclei at the extremes. The Experimental Nuclear Physics Group performs research to improve our understanding of nuclei, nuclear reactions, nuclear decay processes and nuclear astrophysics; an expertise utilized for important laboratory national security programs and for world-class peer-reviewed basic research.

  16. 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").

  17. 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.

  18. 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)

  19. 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.

  20. Modelling oxidation behaviour in operating defective nuclear reactor fuel elements

    NASA Astrophysics Data System (ADS)

    Higgs, Jamie D.

    CANDU nuclear reactors are powered by ceramic uranium dioxide (UO 2) fuel pellets encased in a zirconium-alloy sheath. Occasionally, holes develop in the sheath, allowing steam ingress into the fuel-to-sheath gap, thus exposing the fuel to an oxidizing environment. Oxidation of UO2 fuel may lead to a reduction of fuel thermal conductivity and melting point, both reducing the margin to prevent fuel centre-line melting during transient or even normal operating conditions. Along with increasing fuel temperature, fuel oxidation also enhances the release of radioactive fission products into the reactor coolant. For the first time, a mechanistic treatment has been considered to predict fuel oxidation behaviour in operating defective fuel elements by coupling fuel oxidation kinetics, interstitial oxygen diffusion and heat transfer with sheath oxidation and hydriding rates and gas phase transport in both the fuel-to-sheath gap and within the fuel cracks. The three highly non-linear phenomena (solid-state oxygen diffusion, gas-phase transport and heat transfer) coupled in this treatment were modelled using a finite element technique. The result is a numerical tool that can provide predictions of both the temperature and oxygen-to-uranium (O/U) ratio profile both radially and axially along the fuel element length. The two-dimensional (azimuthally-symmetric) model has been compared to oxygen profile measurements from commercial reactor defective fuel with operating linear power ratings ranging from 26 to 51 kW m-1. Model predictions agree well with experimental observations. Defect size, linear power rating and post-defect residence time (PDRT) appear to be the factors that most influence the extent and rate of fuel oxidation. Thermodynamic modelling of hyperstoichiometric fuel provided the boundary conditions for the fuel oxidation kinetics model. A refined thermodynamic treatment for hyperstoichiometric UO2 has been established. Neutron diffraction experiments at Los Alamos

  1. 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..

  2. 3D laser inspection of fuel assembly grid spacers for nuclear reactors based on diffractive optical elements

    NASA Astrophysics Data System (ADS)

    Finogenov, L. V.; Lemeshko, Yu A.; Zav'yalov, P. S.; Chugui, Yu V.

    2007-06-01

    Ensuring the safety and high operation reliability of nuclear reactors takes 100% inspection of geometrical parameters of fuel assemblies, which include the grid spacers performed as a cellular structure with fuel elements. The required grid spacer geometry of assembly in the transverse and longitudinal cross sections is extremely important for maintaining the necessary heat regime. A universal method for 3D grid spacer inspection using a diffractive optical element (DOE), which generates as the structural illumination a multiple-ring pattern on the inner surface of a grid spacer cell, is investigated. Using some DOEs one can inspect the nomenclature of all produced grids. A special objective has been developed for forming the inner surface cell image. The problems of diffractive elements synthesis, projecting optics calculation, adjusting methods as well as calibration of the experimental measuring system are considered. The algorithms for image processing for different constructive elements of grids (cell, channel hole, outer grid spacer rim) and the experimental results are presented.

  3. 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

  4. 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.

  5. Structural Safety Analysis Based on Seismic Service Conditions for Butterfly Valves in a Nuclear Power Plant

    PubMed Central

    Han, Sang-Uk; Ahn, Dae-Gyun; Lee, Myeong-Gon

    2014-01-01

    The structural integrity of valves that are used to control cooling waters in the primary coolant loop that prevents boiling within the reactor in a nuclear power plant must be capable of withstanding earthquakes or other dangerous situations. In this study, numerical analyses using a finite element method, that is, static and dynamic analyses according to the rigid or flexible characteristics of the dynamic properties of a 200A butterfly valve, were performed according to the KEPIC MFA. An experimental vibration test was also carried out in order to verify the results from the modal analysis, in which a validated finite element model was obtained via a model-updating method that considers changes in the in situ experimental data. By using a validated finite element model, the equivalent static load under SSE conditions stipulated by the KEPIC MFA gave a stress of 135 MPa that occurred at the connections of the stem and body. A larger stress of 183 MPa was induced when we used a CQC method with a design response spectrum that uses 2% damping ratio. These values were lower than the allowable strength of the materials used for manufacturing the butterfly valve, and, therefore, its structural safety met the KEPIC MFA requirements. PMID:24955416

  6. 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.

  7. 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.

  8. 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.

  9. Chemical and nuclear properties of lawrencium (element 103) and hahnium (element 105)

    SciTech Connect

    Henderson, R.A.

    1990-09-10

    The chemical and nuclear properties of Lr and Ha have been studied, using 3-minute {sup 260}Lr and 35-second {sup 262}Ha. The crystal ionic radius of Lr{sup 3+} was determined by comparing its elution position from a cation-exchange resin column with those of lanthanide elements having known ionic radii. Comparisons are made to the ionic radii of the heavy actinides, Am{sup 3+} through Es{sup 3+}, obtained by x-ray diffraction methods, and to Md{sup 3+} and Fm{sup 3+} which were determined in the same manner as Lr{sup 3+}. The hydration enthalpy of {minus}3622 kJ/mol was calculated from the crystal ionic radius using an empirical form of the Born equation. Comparisons to the spacings between the ionic radii of the heaviest members of the lanthanide series show that the 2Z spacing between Lr{sup 3+} and Md{sup 3+} is anomalously small, as the ionic radius of Lr{sup 3+} of 0.0886 nm is significantly smaller than had been expected. The chemical properties of Ha were determined relative to the lighter homologs in group 5, Nb and Ta. Group 4 and group 5 tracer activities, as well as Ha, were absorbed onto glass surfaces as a first step toward the determination of the chemical properties of Ha. Ha was found to adsorb on surfaces, a chemical property unique to the group 5 elements, and as such demonstrates that Ha has the chemical properties of a group 5 element. A solvent extraction procedure was adapted for use as a micro-scale chemical procedure to examine whether or not Ha displays eka-Ta-like chemical under conditions where Ta will be extracted into the organic phase and Nb will not. Under the conditions of this experiment Ha did not extract, and does not show eka-Ta-like chemical properties.

  10. Numerical analysis of a nuclear fuel element for nuclear thermal propulsion

    NASA Technical Reports Server (NTRS)

    Wang, Ten-See; Schutzenhofer, Luke

    1991-01-01

    A computational fluid dynamics model with porosity and permeability formulations in the transport equations has been developed to study the concept of nuclear thermal propulsion through the analysis of a pulsed irradiation of a particle bed element (PIPE). The numerical model is a time-accurate pressure-based formulation. An adaptive upwind scheme is employed for spatial discretization. The upwind scheme is based on second- and fourth-order central differencing with adaptive artificial dissipation. Multiblocked porosity regions have been formulated to model the cold frit, particle bed, and hot frit. Multiblocked permeability regions have been formulated to describe the flow shaping effect from the thickness-varying cold frit. Computational results for several zero-power density PIPEs and an elevated-particle-temperature PIPE are presented. The implications of the computational results are discussed.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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...

  18. 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.

  19. 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.

  20. 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...

  1. 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:...

  2. 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.

  3. Initial Operation and Shakedown of the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Technical Reports Server (NTRS)

    Emrich, William J., Jr.

    2014-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Prototypical fuel elements mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission in addition to being exposed to flowing hydrogen. Recent upgrades to NTREES now allow power levels 24 times greater than those achievable in the previous facility configuration. This higher power operation will allow near prototypical power densities and flows to finally be achieved in most prototypical fuel elements.

  4. NUCLEAR CHEMISTRY: Element 107 Leaves the Table Unturned.

    PubMed

    Service, R F

    2000-08-25

    This week, an international team of chemists reported on the first successful analysis of the chemical properties of bohrium, element 107. The results matched predictions, postponing scientists' hopes of seeing interesting deviations from theory among ultraheavy elements.

  5. 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.

  6. 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.

  7. 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.

  8. 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)

  9. 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.

  10. Prototype Input and Output Data Elements for the Occupational Health and Safety Information System

    NASA Technical Reports Server (NTRS)

    Whyte, A. A.

    1980-01-01

    The National Aeronautics and Space Administration plans to implement a NASA-wide computerized information system for occupational health and safety. The system is necessary to administer the occupational health and safety programs and to meet the legal and regulatory reporting, recordkeeping, and surveillance requirements. Some of the potential data elements that NASA will require as input and output for the new occupational health and safety information system are illustrated. The data elements are shown on sample forms that have been compiled from various sources, including NASA Centers and industry.

  11. 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.

  12. 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.

  13. Status and future of nuclear matrix elements for neutrinoless double-beta decay: a review

    NASA Astrophysics Data System (ADS)

    Engel, Jonathan; Menéndez, Javier

    2017-04-01

    The nuclear matrix elements that govern the rate of neutrinoless double beta decay must be accurately calculated if experiments are to reach their full potential. Theorists have been working on the problem for a long time but have recently stepped up their efforts as ton-scale experiments have begun to look feasible. Here we review past and recent work on the matrix elements in a wide variety of nuclear models and discuss work that will be done in the near future. Ab initio nuclear-structure theory, which is developing rapidly, holds out hope of more accurate matrix elements with quantifiable error bars.

  14. Status and future of nuclear matrix elements for neutrinoless double-beta decay: a review.

    PubMed

    Engel, Jonathan; Menéndez, Javier

    2017-04-01

    The nuclear matrix elements that govern the rate of neutrinoless double beta decay must be accurately calculated if experiments are to reach their full potential. Theorists have been working on the problem for a long time but have recently stepped up their efforts as ton-scale experiments have begun to look feasible. Here we review past and recent work on the matrix elements in a wide variety of nuclear models and discuss work that will be done in the near future. Ab initio nuclear-structure theory, which is developing rapidly, holds out hope of more accurate matrix elements with quantifiable error bars.

  15. Nuclear microscopy in trace-element biology — from cellular studies to the clinic

    NASA Astrophysics Data System (ADS)

    Lindh, Ulf

    1993-05-01

    The concentration and distribution of trace and major elements in cells are of great interest in cell biology. PIXE can provide elemental concentrations in the bulk of cells or organelles as other bulk techniques such as atomic absorption spectrophotometry and nuclear activation analysis. Supplementary information, perhaps more exciting, on the intracellular distributions of trace elements can be provided using nuclear microscopy. Intracellular distributions of trace elements in normal and malignant cells are presented. The toxicity of mercury and cadmium can be prevented by supplementation of the essential trace element selenium. Some results from an experimental animal model are discussed. The intercellular distribution of major and trace elements in isolated blood cells, as revealed by nuclear microscopy, provides useful clinical information. Examples are given concerning inflammatory connective-tissue diseases and the chronic fatigue syndrome.

  16. Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

    SciTech Connect

    Montierth, Leland M.

    2016-07-19

    The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

  17. Nuclear matrix elements for double-β decay

    SciTech Connect

    Engel, Jonathan

    2015-07-15

    Recent progress in nuclear-structure theory has been dramatic. I describe applications in progress of ab inito calculations to double-beta decay, and discuss the recent and future application of generator-coordinate methods to the same problem. I also discuss the old and vexing problem of the renormalization of the weak nuclear axial-vector coupling constant “in medium” and plans to resolve it.

  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. Trace element measurement for assessment of dog food safety.

    PubMed

    De Nadai Fernandes, Elisabete A; Elias, Camila; Bacchi, Márcio Arruda; Bode, Peter

    2017-02-12

    The quality of dog diets depends on adequate ingredients capable of providing optimal nutrition and free of contaminants, for promoting long-term health. Trace elements in 95 samples of dry food for dog puppies (n = 32) and adults (n = 63) of various brands were measured using instrumental neutron activation analysis (INAA). The mass fractions of most elements were within the permissible limits for dogs. Aluminum, antimony, and uranium presented fairly high levels in some samples, which may imply health risks. Aluminum mass fractions ranged from <21 to 11,900 mg/kg, in same brand, super-premium dog food. Antimony mass fractions ranged up to 5.14 mg/kg, with the highest values measured in six samples of dog food from the same producer. The mass fractions of uranium was found up to 4 mg/kg in commercial brands from five different producers.

  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. PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

    DOEpatents

    Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

    1963-09-01

    A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

  4. FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Abbott, W.E.; Balent, R.

    1958-09-16

    A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

  5. Nuclear breeder reactor fuel element with silicon carbide getter

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1987-01-01

    An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

  6. The Role of Nuclear Physics in Understanding the Cosmos and the Origin of Elements

    SciTech Connect

    Balantekin, A. B.

    2011-05-06

    This popular lecture, given in the conference celebrating contributions of Akito Arima to physics on the occasion of his 80th anniversary, outlines the role of nuclear physics in understanding the origin of elements.

  7. Conceptual Software Reliability Prediction Models for Nuclear Power Plant Safety Systems

    SciTech Connect

    Johnson, G.; Lawrence, D.; Yu, H.

    2000-04-03

    The objective of this project is to develop a method to predict the potential reliability of software to be used in a digital system instrumentation and control system. The reliability prediction is to make use of existing measures of software reliability such as those described in IEEE Std 982 and 982.2. This prediction must be of sufficient accuracy to provide a value for uncertainty that could be used in a nuclear power plant probabilistic risk assessment (PRA). For the purposes of the project, reliability was defined to be the probability that the digital system will successfully perform its intended safety function (for the distribution of conditions under which it is expected to respond) upon demand with no unintended functions that might affect system safety. The ultimate objective is to use the identified measures to develop a method for predicting the potential quantitative reliability of a digital system. The reliability prediction models proposed in this report are conceptual in nature. That is, possible prediction techniques are proposed and trial models are built, but in order to become a useful tool for predicting reliability, the models must be tested, modified according to the results, and validated. Using methods outlined by this project, models could be constructed to develop reliability estimates for elements of software systems. This would require careful review and refinement of the models, development of model parameters from actual experience data or expert elicitation, and careful validation. By combining these reliability estimates (generated from the validated models for the constituent parts) in structural software models, the reliability of the software system could then be predicted. Modeling digital system reliability will also require that methods be developed for combining reliability estimates for hardware and software. System structural models must also be developed in order to predict system reliability based upon the reliability

  8. 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.

  9. Advanced Ceramics for Use as Fuel Element Materials in Nuclear Thermal Propulsion Systems

    NASA Technical Reports Server (NTRS)

    Valentine, Peter G.; Allen, Lee R.; Shapiro, Alan P.

    2012-01-01

    With the recent start (October 2011) of the joint National Aeronautics and Space Administration (NASA) and Department of Energy (DOE) Advanced Exploration Systems (AES) Nuclear Cryogenic Propulsion Stage (NCPS) Program, there is renewed interest in developing advanced ceramics for use as fuel element materials in nuclear thermal propulsion (NTP) systems. Three classes of fuel element materials are being considered under the NCPS Program: (a) graphite composites - consisting of coated graphite elements containing uranium carbide (or mixed carbide), (b) cermets (ceramic/metallic composites) - consisting of refractory metal elements containing uranium oxide, and (c) advanced carbides consisting of ceramic elements fabricated from uranium carbide and one or more refractory metal carbides [1]. The current development effort aims to advance the technology originally developed and demonstrated under Project Rover (1955-1973) for the NERVA (Nuclear Engine for Rocket Vehicle Application) [2].

  10. 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)

  11. NUCLEAR REACTOR FUEL ELEMENT AND METHOD OF MANUFACTURE

    DOEpatents

    Brooks, H.

    1960-04-26

    A description is given for a fuel element comprising a body of uranium metal or an uranium compound dispersed in a matrix material made from magnesium, calcium, or barium and a stainless steel jacket enclosing the body.

  12. Technology Status of Thermionic Fuel Elements for Space Nuclear Power

    NASA Technical Reports Server (NTRS)

    Holland, J. W.; Yang, L.

    1984-01-01

    Thermionic reactor power systems are discussed with respect to their suitability for space missions. The technology status of thermionic emitters and sheath insulator assemblies is described along with testing of the thermionic fuel elements.

  13. 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.

  14. 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.

  15. NUCLEAR REACTOR FUEL ELEMENTS AND METHOD OF PREPARATION

    DOEpatents

    Kingston, W.E.; Kopelman, B.; Hausner, H.H.

    1963-07-01

    A fuel element consisting of uranium nitride and uranium carbide in the form of discrete particles in a solid coherent matrix of a metal such as steel, beryllium, uranium, or zirconium and clad with a metal such as steel, aluminum, zirconium, or beryllium is described. The element is made by mixing powdered uranium nitride and uranium carbide with powdered matrix metal, then compacting and sintering the mixture. (AEC)

  16. Dynamic SPR monitoring of yeast nuclear protein binding to a cis-regulatory element

    SciTech Connect

    Mao, Grace; Brody, James P.

    2007-11-09

    Gene expression is controlled by protein complexes binding to short specific sequences of DNA, called cis-regulatory elements. Expression of most eukaryotic genes is controlled by dozens of these elements. Comprehensive identification and monitoring of these elements is a major goal of genomics. In pursuit of this goal, we are developing a surface plasmon resonance (SPR) based assay to identify and monitor cis-regulatory elements. To test whether we could reliably monitor protein binding to a regulatory element, we immobilized a 16 bp region of Saccharomyces cerevisiae chromosome 5 onto a gold surface. This 16 bp region of DNA is known to bind several proteins and thought to control expression of the gene RNR1, which varies through the cell cycle. We synchronized yeast cell cultures, and then sampled these cultures at a regular interval. These samples were processed to purify nuclear lysate, which was then exposed to the sensor. We found that nuclear protein binds this particular element of DNA at a significantly higher rate (as compared to unsynchronized cells) during G1 phase. Other time points show levels of DNA-nuclear protein binding similar to the unsynchronized control. We also measured the apparent association complex of the binding to be 0.014 s{sup -1}. We conclude that (1) SPR-based assays can monitor DNA-nuclear protein binding and that (2) for this particular cis-regulatory element, maximum DNA-nuclear protein binding occurs during G1 phase.

  17. 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

  18. 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

  19. 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).

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Applications of a global nuclear-structure model to studies of the heaviest elements

    SciTech Connect

    Moeller, P.; Nix, J.R.

    1993-10-01

    We present some new results on heavy-element nuclear-structure properties calculated on the basis of the finite-range droplet model and folded-Yukawa single-particle potential. Specifically, we discuss calculations of nuclear ground-state masses and microscopic corrections, {alpha}-decay properties, {beta}-decay properties, fission potential-energy surfaces, and spontaneous-fission half-lives. These results, obtained in a global nuclear-structure approach, are particularly reliable for describing the stability properties of the heaviest elements.

  13. 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.

  14. 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.

  15. Molten tin reprocessing of spent nuclear fuel elements

    DOEpatents

    Heckman, Richard A.

    1983-01-01

    A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

  16. 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.

  17. Support grid for fuel elements in a nuclear reactor

    DOEpatents

    Finch, Lester M.

    1977-01-01

    A support grid is provided for holding nuclear fuel rods in a rectangular array. Intersecting sheet metal strips are interconnected using opposing slots in the strips to form a rectangular cellular grid structure for engaging the sides of a multiplicity of fuel rods. Spring and dimple supports for engaging fuel and guide rods extending through each cell in the support grid are formed in the metal strips with the springs thus formed being characterized by nonlinear spring rates.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements

    SciTech Connect

    Leland M. Montierth

    2010-12-01

    The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

  3. The Search for Transuranium Elements and the Discovery of Nuclear Fission

    NASA Astrophysics Data System (ADS)

    Sime, Ruth Lewin

    The synthesis of new, artificial elements beyond uranium was at the cutting-edge of physical research in the 1930s, and nearly half a dozen transuranium elements were reported between 1934 and 1938. Nuclear physicists and radiochemists collaborated closely, but each field introduced fundamental assumptions that proved to be false: that nuclear changes would always be small, and that transuranium elements would resemble transition elements chemically. With its surprise ending in the discovery of nuclear fission, the misguided transuranium project can be viewed as an example of the illogical progress of scientific discovery. It is also an example of an interdisciplinary collaboration that was flawed yet crucial, for although chemists and physicists both contributed to the delay in discovering fission, their collaboration was essential in leading them to it in the end.

  4. Recovery of transplutonium elements from nuclear reactor waste

    DOEpatents

    Campbell, David O.; Buxton, Samuel R.

    1977-05-24

    A method of separating actinide values from nitric acid waste solutions resulting from reprocessing of irradiated nuclear fuels comprises oxalate precipitation of the major portion of actinide and lanthanide values to provide a trivalent fraction suitable for subsequent actinide/lanthanide partition, exchange of actinide and lanthanide values in the supernate onto a suitable cation exchange resin to provide an intermediate-lived raffinate waste stream substantially free of actinides, and elution of the actinide values from the exchange resin. The eluate is then used to dissolve the trivalent oxalate fraction prior to actinide/lanthanide partition or may be combined with the reprocessing waste stream and recycled.

  5. 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.

  6. 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.

  7. 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.

  8. METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

    DOEpatents

    Handwerk, J.H.; BAch, R.A.

    1959-08-18

    A method is described for preparing a reactor fuel element by forming a mixture of thorium dioxide and an oxide of uranium, the uranium being present. In an oxidation state at least as high as it is in U/sub 3/O/sub 8/, into a desired shape and firing in air at a temperature siifficiently high to reduce the higher uranium oxide to uranium dioxide.

  9. FUEL ELEMENTS FOR NUCLEAR REACTORS AND PROCESS OF MAKING

    DOEpatents

    Roake, W.E.

    1958-08-19

    A process is described for producing uranium metal granules for use in reactor fuel elements. The granules are made by suspending powdered uramiunn metal or uranium hydride in a viscous, non-reactive liquid, such as paraffin oil, aad pouring the resulting suspension in droplet, on to a bed of powdered absorbent. In this manner the liquid vehicle is taken up by the sorbent and spherical pellets of uranium metal are obtained. The

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. Multi-Element CZT Array for Nuclear Safeguards Applications

    NASA Astrophysics Data System (ADS)

    Kwak, S.-W.; Lee, A.-R.; Shin, J.-K.; Park, U.-R.; Park, S.; Kim, Y.; Chung, H.

    2016-12-01

    Due to its electronic properties, a cadmium zinc telluride (CZT) detector has been used as a hand-held portable nuclear measurement instrument. However, a CZT detector has low detection efficiency because of a limitation of its single crystal growth. To address its low efficiency, we have constructed a portable four-CZT array based gamma-ray spectrometer consisting of a CZT array, electronics for signal processing and software. Its performance has been characterized in terms of energy resolution and detection efficiency using radioactive sources and nuclear materials. Experimental results showed that the detection efficiency of the four-CZT array based gamma-ray spectrometer was much higher than that of a single CZT detector in the array. The FWHMs of the CZT array were 9, 18, and 21 keV at 185.7, 662, and 1,332 keV, respectively. Some gamma-rays in a range of 100 keV to 200 keV were not clear in a single crystal detector while those from the CZT array system were observed to be clear. The energy resolution of the CZT array system was only slightely worse than those of the single CZT detectors. By combining several single crystals and summing signals from each single detector at a digital electronic circuit, the detection efficiency of a CZT array system increased without degradation of its energy resolution. The technique outlined in this paper shows a very promising method for designing a CZT-based gamma-ray spectroscopy that overcomes the fundamental limitations of a small volume CZT detector.

  17. 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)

  18. 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…

  19. 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.

  20. 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...

  1. 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.

  2. Allocating resources and building confidence in public-safety decisions for nuclear waste sites

    SciTech Connect

    Lew, K L; Wilder, D G

    1999-05-21

    There are three basic ways to protect the public from the hazards of exposure to radionuclides in nuclear waste: completely contain the waste; limit the rate at which radionuclides are released; and, once radionuclides are released, minimize their impact by reducing concentrations and retarding transport. A geologic repository system that implements all three provides maximum protection for the public: if one element fails, the others serve to protect. This is ''defense-in-depth.'' Demonstrating confidence in the ability of a designed system to provide the requisite safety to the public must rely on a combination of the following aspects relating to engineered and natural system components: 1 Knowledge or understanding of properties and processes 2 Uniformity of (or ability to understand or control) the range of variability associated with each component 3 Experience over time This paper proposes a tool based on defining a ''confidence region'' determined by these three essential aspects of confidence. The defense-in-depth decision-making tool described identifies the portion of the ultimate confidence region that is not well demonstrated and indicates where there is potential for changing a specific component's confidence region, therefore providing in-formation for decisions on emphasis--either for demonstrating performance or for focusing on further studies. The US Yucca Mountain Site Characterization Project (YMP), wherein Yucca Mountain is being investigated as a potential site for a nuclear waste repository, and the Swedish geologic repository studies are used as examples of this tool. of protective or operating components such that failure of a single component does not by itself lead to system failure. The greater the exposure to loss, the greater the requirements for design margins (the margin of conservatism associated with the fabrication and operation of important components in complex engineering projects) or for compensation by defense-in-depth. Thus

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Preliminary study of nuclear fuel element testing based on coded source neutron imaging

    SciTech Connect

    Sheng Wang; Hang Li; Chao Cao; Yang Wu; Heyong Huo; Bin Tang

    2015-07-01

    Neutron radiography (NR) is one of the most important nondestructive testing methods, which is sensitive to low density materials. Especially, Neutron transfer imaging method could be used to test radioactivity materials refraining from γ effect, but it is difficult to realize tomography. Coded source neutron imaging (CSNI) is a newly NR method developed fast in the last several years. The distance between object and detector is much longer than traditional NR, which could be used to test radioactivity materials. With pre-reconstruction process from fold-cover projections, CSNI could easily realize tomography. This thesis carries out preliminary study on the nuclear fuel element testing by coded source neutron imaging. We calculate different enrichment, flaws and activity in nuclear fuel elements tested by CSNI with Monte-Carlo simulation. The results show that CSNI could be a useful testing method for nuclear fuel element testing. (authors)

  10. Chemical and nuclear properties of Rutherfordium (Element 104)

    SciTech Connect

    Kacher, Christian D.

    1995-10-30

    The chemical-properties of rutherfordium (Rf) and its group 4 homologs were studied by sorption on glass support surfaces coated with cobalt(II)ferrocyanide and by solvent extraction with tributylphosphate (TBP) and triisooctylamine (TIOA). The surface studies showed that the hydrolysis trend in the group 4 elements and the pseudogroup 4 element, lb, decreases in the order Rf>Zr≈Hf>Th. This trend was attributed to relativistic effects which predicted that Rf would be more prone to having a coordination number of 6 than 8 in most aqueous solutions due to a destabilization of the 6d5/2 shell and a stabilization of the 7pI/2 shell. This hydrolysis trend was confirmed in the TBP/HBr solvent extraction studies which showed that the extraction trend decreased in the order Zr>Hf>Rf?Ti for HBr, showing that Rf and Ti did not extract as well because they hydrolyzed more easily than Zr and Hf. The TIOA/HF solvent extraction studies showed that the extraction trend for the group 4 elements decreased in the order Ti>Zr≈Hf>Rf, in inverse order from the trend of ionic radii Rf>Zr≈Hf>Ti. An attempt was made to produce 263Rf (a) via the 248Cm(22Ne, α3n) reaction employing thenoyltrifluoroacetone (TTA) solvent extraction chemistry and (b) via the 249Bk(18O,4n) reaction employing the Automated Rapid Chemistry Apparatus (ARCA). In the TTA studies, 16 fissions were observed but were all attributed to 256Fm. No alpha events were observed in the Rf chemical fraction. A 0.2 nb upper limit production cross section for the 248Cm(22Ne, α3n)263Rf reaction was calculated assuming the 500-sec half-life reported previously by Czerwinski et al. [CZE92A].

  11. 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.

  12. 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

  13. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

    NASA Technical Reports Server (NTRS)

    Emrich, William J.; Moran, Robert P.; Pearson, J. Bose

    2013-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities. Keywords: Nuclear Thermal Propulsion, Simulator

  14. Double β-decay nuclear matrix elements for the A=48 and A=58 systems

    NASA Astrophysics Data System (ADS)

    Skouras, L. D.; Vergados, J. D.

    1983-11-01

    The nuclear matrix elements entering the double β decays of the 48Ca-48Ti and 58Ni-58Fe systems have been calculated using a realistic two nucleon interaction and realistic shell model spaces. Effective transition operators corresponding to a variety of gauge theory models have been considered. The stability of such matrix elements against variations of the nuclear parameters is examined. Appropriate lepton violating parameters are extracted from the A=48 data and predictions are made for the lifetimes of the positron decays of the A=58 system. RADIOACTIVITY Double β decay. Gauge theories. Lepton nonconservation. Neutrino mass. Shell model calculations.

  15. LDRD Final Report: Surrogate Nuclear Reactions and the Origin of the Heavy Elements (04-ERD-057)

    SciTech Connect

    Escher, J E; Bernstein, L A; Bleuel, D; Burke, J; Church, J A; Dietrich, F S; Forssen, C; Gueorguiev, V; Hoffman, R D

    2007-02-23

    Research carried out in the framework of the LDRD project ''Surrogate Nuclear Reactions and the Origin of the Heavy Elements'' (04-ERD-057) is summarized. The project was designed to address the challenge of determining cross sections for nuclear reactions involving unstable targets, with a particular emphasis on reactions that play a key role in the production of the elements between Iron and Uranium. This report reviews the motivation for the research, introduces the approach employed to address the problem, and summarizes the resulting scientific insights, technical findings, and related accomplishments.

  16. Oxidation state of multivalent elements in high-level nuclear waste glass

    SciTech Connect

    Reynolds, J.G.

    2007-07-01

    Nuclear waste contains many different elements that have more than one oxidation state. When the nuclear waste is treated by vitrification, the behavior of the element in the melter and resulting glass product depends on the stable oxidation state. The stable oxidation state in any medium can be calculated from the standard potential in that medium. Consequently, the standard potential of multi-valent elements has been measured in many silicate-melts, including ones relevant to nuclear waste treatment. In this study, the relationship between the standard potential in molten nuclear waste glass and the standard potential in water will be quantified so that the standard potential of elements that have not been measured in glass can be estimated. The regression equation was found to have an R{sup 2} statistic of 0.96 or 0.83 depending on the number of electrons transferred in the reaction. The Nernst equation was then used to calculate the oxidation state of other relevant multi-valent elements in nuclear waste glass from these standard potentials and the measured ferrous to ferric iron ratio. The calculated oxidation states were consistent with all oxidation state measurements available. The calculated oxidation states were used to rationalize the behavior of many of the multi-valent elements. For instance, chromium increases glass crystallization because it is in the trivalent-state, iodine volatilises from the melter because it is in the volatile zero-valent state, and the leaching behavior of arsenic is driven by its oxidation state. Thus, these thermodynamic calculations explain the behavior of many trace elements during the vitrification process. (authors)

  17. 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...

  18. 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,...

  19. Recent studies of nuclear and chemical properties of elements 103, 104 and 105

    SciTech Connect

    Hoffman, D.C. . Dept. of Chemistry Lawrence Berkeley Lab., CA )

    1990-08-01

    Information obtained since 1983 on the nuclear and chemical properties of element 103, the last on the actinide series, and elements 104 and 105, at the beginning of the transactinide series, is reviewed. Their chemical properties are compared with their lanthanide and lighter group 4 and 5 homologs and evidence for possible relativistic effects is discussed. The current knowledge of the nuclear properties of these elements and how these affect of the study of chemical properties is discussed. Some of the challenges involved in the study of short-lived isotopes which can only be produced an atom-at-a-time'' at an appropriate accelerator and the prognosis for future studies of these and still heavier elements are considered. 40 refs., 4 figs.

  20. Recapturing Graphite-Based Fuel Element Technology for Nuclear Thermal Propulsion

    SciTech Connect

    Trammell, Michael P; Jolly, Brian C; Miller, James Henry; Qualls, A L; Harrison, Thomas J

    2013-01-01

    ORNL is currently recapturing graphite based fuel forms for Nuclear Thermal Propulsion (NTP). This effort involves research and development on materials selection, extrusion, and coating processes to produce fuel elements representative of historical ROVER and NERVA fuel. Initially, lab scale specimens were fabricated using surrogate oxides to develop processing parameters that could be applied to full length NTP fuel elements. Progress toward understanding the effect of these processing parameters on surrogate fuel microstructure is presented.

  1. 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.

  2. 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.

  3. Nuclear matrix element of neutrinoless double-β decay: Relativity and short-range correlations

    NASA Astrophysics Data System (ADS)

    Song, L. S.; Yao, J. M.; Ring, P.; Meng, J.

    2017-02-01

    Background:The discovery of neutrinoless double-β (0 ν β β ) decay would demonstrate the nature of neutrinos, have profound implications for our understanding of matter-antimatter mystery, and solve the mass hierarchy problem of neutrinos. The calculations for the nuclear matrix elements M0 ν of 0 ν β β decay are crucial for the interpretation of this process. Purpose: We study the effects of relativity and nucleon-nucleon short-range correlations on the nuclear matrix elements M0 ν by assuming the mechanism of exchanging light or heavy neutrinos for the 0 ν β β decay. Methods:The nuclear matrix elements M0 ν are calculated within the framework of covariant density functional theory, where the beyond-mean-field correlations are included in the nuclear wave functions by configuration mixing of both angular-momentum and particle-number projected quadrupole deformed mean-field states. Results: The nuclear matrix elements M0 ν are obtained for ten 0 ν β β -decay candidate nuclei. The impact of relativity is illustrated by adopting relativistic or nonrelativistic decay operators. The effects of short-range correlations are evaluated. Conclusions: The effects of relativity and short-range correlations play an important role in the mechanism of exchanging heavy neutrinos though the influences are marginal for light neutrinos. Combining the nuclear matrix elements M0 ν with the observed lower limits on the 0 ν β β -decay half-lives, the predicted strongest limits on the effective masses are ||<0.06 eV for light neutrinos and | |-1>3.065 ×108GeV for heavy neutrinos.

  4. Multiphysics Modeling of a Single Channel in a Nuclear Thermal Propulsion Grooved Ring Fuel Element

    NASA Technical Reports Server (NTRS)

    Kim, Tony; Emrich, William J., Jr.; Barkett, Laura A.; Mathias, Adam D.; Cassibry, Jason T.

    2013-01-01

    In the past, fuel rods have been used in nuclear propulsion applications. A new fuel element concept that reduces weight and increases efficiency uses a stack of grooved discs. Each fuel element is a flat disc with a hole on the interior and grooves across the top. Many grooved ring fuel elements for use in nuclear thermal propulsion systems have been modeled, and a single flow channel for each design has been analyzed. For increased efficiency, a fuel element with a higher surface-area-to-volume ratio is ideal. When grooves are shallower, i.e., they have a lower surface area, the results show that the exit temperature is higher. By coupling the physics of turbulence with those of heat transfer, the effects on the cooler gas flowing through the grooves of the thermally excited solid can be predicted. Parametric studies were done to show how a pressure drop across the axial length of the channels will affect the exit temperatures of the gas. Geometric optimization was done to show the behaviors that result from the manipulation of various parameters. Temperature profiles of the solid and gas showed that more structural optimization is needed to produce the desired results. Keywords: Nuclear Thermal Propulsion, Fuel Element, Heat Transfer, Computational Fluid Dynamics, Coupled Physics Computations, Finite Element Analysis

  5. 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.

  6. 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

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. An assessment of swaged connections in a nuclear fuel element using nonlinear finite element analysis

    SciTech Connect

    Richins, W.D.; Miller, G.K.

    1995-12-01

    Large displacement, non-linear finite element analyses were performed to evaluate a swaging process used to fabricate connections between plates in the fuel elements for a test reactor at the Idaho National Engineering Laboratory. The force required to pull the fuel plate from the connection is referred to as the strength of the connection. Assurance that the integrity of the connections is maintained through reactor operation is provided by establishing a minimum acceptance requirement for this strength. Analysis results were used to assess the sensitivity of the strength of the swaged connections to variations in several manufacturing process parameters. The predicted strengths correlated well with results from tests where sample swaged connections were loaded to failure. Results from these investigations were used to assess the adequacy and need for various fabrication, testing, and quality control requirements.

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. COATED CARBON ELEMENT FOR USE IN NUCLEAR REACTORS AND THE PROCESS OF MAKING THE ELEMENT

    DOEpatents

    Pyle, R.J.; Allen, G.L.

    1963-01-15

    S>This patent relates to a carbide-nitride-carbide coating for carbon bodies that are to be subjected to a high temperature nuclear reactor atmosphere, and a method of applying the same. This coating is a highly efficient diffusion barrier and protects the C body from corrosion and erosion by the reactor atmosphere. Preferably, the innermost coating is Zr carbide, the middle coatlng is Zr nitride, and the outermost coating is a mixture of Zr and Nb carbide. The nitride coating acts as a diffusion barrier, while the innermost carbide bonds the nitride to the C body and prevents deleterious reaction between the nitride and C body. The outermost carbide coating protects the nitride coating from the reactor atmosphere. (AEC)

  17. Role of nuclear analytical probe techniques in biological trace element research

    SciTech Connect

    Jones, K.W.; Pounds, J.G.

    1985-01-01

    Many biomedical experiments require the qualitative and quantitative localization of trace elements with high sensitivity and good spatial resolution. The feasibility of measuring the chemical form of the elements, the time course of trace elements metabolism, and of conducting experiments in living biological systems are also important requirements for biological trace element research. Nuclear analytical techniques that employ ion or photon beams have grown in importance in the past decade and have led to several new experimental approaches. Some of the important features of these methods are reviewed here along with their role in trace element research, and examples of their use are given to illustrate potential for new research directions. It is emphasized that the effective application of these methods necessitates a closely integrated multidisciplinary scientific team. 21 refs., 4 figs., 1 tab.

  18. 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.

  19. Association of Nuclear Localization of a Long Interspersed Nuclear Element-1 Protein in Breast Tumors with Poor Prognostic Outcomes

    PubMed Central

    Harris, Chris R.; Normart, Robin; Yang, Qifeng; Stevenson, Elizabeth; Haffty, Bruce G.; Ganesan, Shridar; Cordon-Cardo, Carlos; Levine, Arnold J.; Tang, Laura H.

    2010-01-01

    Within healthy human somatic cells, retrotransposition by long interspersed nuclear element-1 (also known as LINE-1 or L1) is thought to be held in check by a variety of mechanisms, including DNA methylation and RNAi. The expression of L1-ORF1 protein, which is rarely found in normal tissue, was assayed using antibodies with a variety of clinical cancer specimens and cancer cell lines. L1-ORF1p expression was detected in nearly all breast tumors that the authors examined, and the protein was also present in a high percentage of ileal carcinoids, bladder, and pancreatic neuroendocrine tumors, as well as in a smaller percentage of prostate and colorectal tumors. Tumors generally demonstrated cytoplasmic L1-ORF1p; however, in several breast cancers, L1-ORF1p was nuclear. Patients with breast tumors displaying nuclear L1-ORF1p had a greater incidence of both local recurrence and distal metastases and also showed poorer overall survival when compared with patients with tumors displaying cytoplasmic L1-ORF1p. These data suggest that expression of L1-ORF1p is widespread in many cancers and that redistribution from cytoplasm to nucleus could be a poor prognostic indicator during breast cancer. High expression and nuclear localization of L1-ORF1p may result in a higher rate of L1 retrotransposition, which could increase genomic instability. PMID:20948976

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. Preparation of Metal Filter Element for Fail Safety in IGCC Filter Unit

    SciTech Connect

    Choi, J-H.; Ahn, I-S.; Bak, Y-C.; Bae, S-Y.; Ha, S-J.; Jang, H-J.

    2002-09-18

    Metal filter elements as the fail safety filter are fabricated by the methods using cold isostatic pressure (compress method) and binder (binder method) to form the filter element and tested in a experimental and bench units. The fail safety filter on the filtration system is mounted additionally in order to intercept the particle leak when the main filter element is broken. So it should have two contrary functions of a high permeability and being plugged easily. The filter element having high porosity and high plugging property was fabricated by the bind method. It has the porosity more than 50%, showed very small pressure drop less than 10mmH2O at the face velocity of 0.15m/s, and plugged within 5 minutes with the inhibition of the particle leak larger than 4 {micro}m. The test result of corrosion tendency in IGCC gas stream at 500 C shows SUS310L material is very reasonable among SUS310, SUS316, Inconel 600, and Hastelloy X.

  5. Modeling and Testing of Non-Nuclear, Highpower Simulated Nuclear Thermal Rocket Reactor Elements

    NASA Technical Reports Server (NTRS)

    Kirk, Daniel R.

    2005-01-01

    When the President offered his new vision for space exploration in January of 2004, he said, "Our third goal is to return to the moon by 2020, as the launching point for missions beyond," and, "With the experience and knowledge gained on the moon, we will then be ready to take the next steps of space exploration: human missions to Mars and to worlds beyond." A human mission to Mars implies the need to move large payloads as rapidly as possible, in an efficient and cost-effective manner. Furthermore, with the scientific advancements possible with Project Prometheus and its Jupiter Icy Moons Orbiter (JIMO), (these use electric propulsion), there is a renewed interest in deep space exploration propulsion systems. According to many mission analyses, nuclear thermal propulsion (NTP), with its relatively high thrust and high specific impulse, is a serious candidate for such missions. Nuclear rockets utilize fission energy to heat a reactor core to very high temperatures. Hydrogen gas flowing through the core then becomes superheated and exits the engine at very high exhaust velocities. The combination of temperature and low molecular weight results in an engine with specific impulses above 900 seconds. This is almost twice the performance of the LOX/LH2 space shuttle engines, and the impact of this performance would be to reduce the trip time of a manned Mars mission from the 2.5 years, possible with chemical engines, to about 12-14 months.

  6. 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.

  7. 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.

  8. 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.

  9. Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques

    SciTech Connect

    Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Diego Mandelli; Michael Pernice; Robert Nourgaliev

    2013-10-01

    A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis

  10. Nuclear astrophysics: the unfinished quest for the origin of the elements

    NASA Astrophysics Data System (ADS)

    José, Jordi; Iliadis, Christian

    2011-09-01

    Half a century has passed since the foundation of nuclear astrophysics. Since then, this discipline has reached its maturity. Today, nuclear astrophysics constitutes a multidisciplinary crucible of knowledge that combines the achievements in theoretical astrophysics, observational astronomy, cosmochemistry and nuclear physics. New tools and developments have revolutionized our understanding of the origin of the elements: supercomputers have provided astrophysicists with the required computational capabilities to study the evolution of stars in a multidimensional framework; the emergence of high-energy astrophysics with space-borne observatories has opened new windows to observe the Universe, from a novel panchromatic perspective; cosmochemists have isolated tiny pieces of stardust embedded in primitive meteorites, giving clues on the processes operating in stars as well as on the way matter condenses to form solids; and nuclear physicists have measured reactions near stellar energies, through the combined efforts using stable and radioactive-ion beam facilities. This review provides comprehensive insight into the nuclear history of the Universe and related topics: starting from the Big Bang, when the ashes from the primordial explosion were transformed to hydrogen, helium and a few trace elements, to the rich variety of nucleosynthesis mechanisms and sites in the Universe. Particular attention is paid to the hydrostatic processes governing the evolution of low-mass stars, red giants and asymptotic giant-branch stars, as well as to the explosive nucleosynthesis occurring in core-collapse and thermonuclear supernovae, γ-ray bursts, classical novae, x-ray bursts, superbursts and stellar mergers.

  11. 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.

  12. 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).

  13. Advanced neutron source reactor conceptual safety analysis report, three-element-core design: Chapter 15, accident analysis

    SciTech Connect

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.; Harrington, R.M.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design for the Advanced Neutron Source has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. To assess the impact of changes in the core region configuration and the thermal-hydraulic steady-state conditions, the safety analysis has been updated. This report gives the safety margins for the loss-of-off-site power and pressure-boundary fault accidents based on the RELAP5 results. AU margins are greater for the three-element-core simulations than those calculated for the two-element core.

  14. Characterization of NiTi Shape Memory Damping Elements designed for Automotive Safety Systems

    NASA Astrophysics Data System (ADS)

    Strittmatter, Joachim; Clipa, Victor; Gheorghita, Viorel; Gümpel, Paul

    2014-07-01

    Actuator elements made of NiTi shape memory material are more and more known in industry because of their unique properties. Due to the martensitic phase change, they can revert to their original shape by heating when subjected to an appropriate treatment. This thermal shape memory effect (SME) can show a significant shape change combined with a considerable force. Therefore such elements can be used to solve many technical tasks in the field of actuating elements and mechatronics and will play an increasing role in the next years, especially within the automotive technology, energy management, power, and mechanical engineering as well as medical technology. Beside this thermal SME, these materials also show a mechanical SME, characterized by a superelastic plateau with reversible elongations in the range of 8%. This behavior is based on the building of stress-induced martensite of loaded austenite material at constant temperature and facilitates a lot of applications especially in the medical field. Both SMEs are attended by energy dissipation during the martensitic phase change. This paper describes the first results obtained on different actuator and superelastic NiTi wires concerning their use as damping elements in automotive safety systems. In a first step, the damping behavior of small NiTi wires up to 0.5 mm diameter was examined at testing speeds varying between 0.1 and 50 mm/s upon an adapted tensile testing machine. In order to realize higher testing speeds, a drop impact testing machine was designed, which allows testing speeds up to 4000 mm/s. After introducing this new type of testing machine, the first results of vertical-shock tests of superelastic and electrically activated actuator wires are presented. The characterization of these high dynamic phase change parameters represents the basis for new applications for shape memory damping elements, especially in automotive safety systems.

  15. Nuclear isomers in superheavy elements as stepping stones towards the island of stability.

    PubMed

    Herzberg, R-D; Greenlees, P T; Butler, P A; Jones, G D; Venhart, M; Darby, I G; Eeckhaudt, S; Eskola, K; Grahn, T; Gray-Jones, C; Hessberger, F P; Jones, P; Julin, R; Juutinen, S; Ketelhut, S; Korten, W; Leino, M; Leppänen, A-P; Moon, S; Nyman, M; Page, R D; Pakarinen, J; Pritchard, A; Rahkila, P; Sarén, J; Scholey, C; Steer, A; Sun, Y; Theisen, Ch; Uusitalo, J

    2006-08-24

    A long-standing prediction of nuclear models is the emergence of a region of long-lived, or even stable, superheavy elements beyond the actinides. These nuclei owe their enhanced stability to closed shells in the structure of both protons and neutrons. However, theoretical approaches to date do not yield consistent predictions of the precise limits of the 'island of stability'; experimental studies are therefore crucial. The bulk of experimental effort so far has been focused on the direct creation of superheavy elements in heavy ion fusion reactions, leading to the production of elements up to proton number Z = 118 (refs 4, 5). Recently, it has become possible to make detailed spectroscopic studies of nuclei beyond fermium (Z = 100), with the aim of understanding the underlying single-particle structure of superheavy elements. Here we report such a study of the nobelium isotope 254No, with 102 protons and 152 neutrons--the heaviest nucleus studied in this manner to date. We find three excited structures, two of which are isomeric (metastable). One of these structures is firmly assigned to a two-proton excitation. These states are highly significant as their location is sensitive to single-particle levels above the gap in shell energies predicted at Z = 114, and thus provide a microscopic benchmark for nuclear models of the superheavy elements.

  16. Nuclear matrix elements of the double beta decay for mass around 80

    NASA Astrophysics Data System (ADS)

    Yoshinaga, Naotaka; Higashiyama, Koji; Teruya, Eri

    2014-09-01

    In nature there are 30 kinds of nuclei which are expected to have double beta decays. Among them ten nuclei are actually observed for the neutrino double beta decays. Still no observation is made for the neutrinoless double beta decays (0 νββ) . The 0 νββ decay is expected to occur only when neutrinos have masses and they are Majorana particles. In that respect observation of 0 νββ is to determine whether neutrinos are Majorana particles or not. In theoretical side in order to estimate the half life of 0 νββ determination of the nuclear matrix elements are essential. They were calculated in many theoretical frameworks, but the results are not consistent in various models. In this study we carry out shell model calculations for 82Se and 82Kr nuclei. After obtaining the wavefunctions, we calculate the nuclear matrix elements. For comparison we make pair truncated shell model calculations.

  17. Influence of Pairing on the Nuclear Matrix Elements of the Neutrinoless {beta}{beta} Decays

    SciTech Connect

    Caurier, E.; Nowacki, F.

    2008-02-08

    We study in this Letter the neutrinoless double beta decay nuclear matrix elements (NME's) in the framework of the interacting shell model. We analyze them in terms of the total angular momentum of the decaying neutron pair and as a function of the seniority truncations in the nuclear wave functions. This point of view turns out to be very adequate to gauge the accuracy of the NME's predicted by different nuclear models. In addition, it gives back the protagonist role in this process to the pairing interaction, the one which is responsible for the very existence of double beta decay emitters. We show that low seniority approximations, comparable to those implicit in the quasiparticle RPA in a spherical basis, tend to overestimate the NME's in several decays.

  18. Design Considerations for the Nuclear Thermal Rocket Element Environmental Simulator (NTREES)

    NASA Technical Reports Server (NTRS)

    Emrich, Bill; Kirk, Daniel

    2006-01-01

    Nuclear Thermal Rockets or NTR's have been suggested as a propulsion system option for vehicles traveling to the moon or Mars. These engines are capable of providing high thrust at specific impulses at least twice that of today s best chemical engines. The performance constraints on these engines are mainly the result of temperature limitations on the fuel coupled with a limited ability to withstand chemical attack by the hot hydrogen propellant. To operate at maximum efficiency, fuel forms are desired which can withstand the extremely hot, hostile environment characteristic of NTR operation for at least several hours. The simulation of such an environment would require an experimental device which could simultaneously approximate the power, flow, and temperature conditions which a nuclear fuel element (or partial element) would encounter during NTR operation. Such a simulation would allow detailed studies of the fuel behavior and hydrogen flow characteristics under reactor like conditions to be performed. The goal of these simulations would be directed toward expanding the performance envelope of NTR engines over that which was demonstrated during the Rover and NERVA nuclear rocket programs of the 1970's. Currently, such a simulator is nearing completion at the Marshall Space Flight Center, and will shortly be used in the future to evaluate a wide variety of he1 element designs and the materials of which they are constructed. This present work addresses the initial experimental objectives of the Nuclear Thermal Rocket Element Environmental Simulator or NTREES and some of the design considerations which were considered prior to and during its construction.

  19. 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...

  20. 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...

  1. 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....

  2. 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....

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. 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.

  8. Nuclear energy.

    PubMed

    Wilson, Peter D

    2010-01-01

    The technical principles and practices of the civil nuclear industry are described with particular reference to fission and its products, natural and artificial radioactivity elements principally concerned and their relationships, main types of reactor, safety issues, the fuel cycle, waste management, issues related to weapon proliferation, environmental considerations and possible future developments.

  9. Relativistic description of nuclear matrix elements in neutrinoless double-β decay

    NASA Astrophysics Data System (ADS)

    Song, L. S.; Yao, J. M.; Ring, P.; Meng, J.

    2014-11-01

    Background: Neutrinoless double-β (0 ν β β ) decay is related to many fundamental concepts in nuclear and particle physics beyond the standard model. Currently there are many experiments searching for this weak process. An accurate knowledge of the nuclear matrix element for the 0 ν β β decay is essential for determining the effective neutrino mass once this process is eventually measured. Purpose: We report the first full relativistic description of the 0 ν β β decay matrix element based on a state-of-the-art nuclear structure model. Methods: We adopt the full relativistic transition operators which are derived with the charge-changing nucleonic currents composed of the vector coupling, axial-vector coupling, pseudoscalar coupling, and weak-magnetism coupling terms. The wave functions for the initial and final nuclei are determined by the multireference covariant density functional theory (MR-CDFT) based on the point-coupling functional PC-PK1. Correlations beyond the mean field are introduced by configuration mixing of both angular momentum and particle number projected quadrupole deformed mean-field wave functions. Results: The low-energy spectra and electric quadrupole transitions in 150Nd and its daughter nucleus 150Sm are well reproduced by the MR-CDFT calculations. The 0 ν β β decay matrix elements for both the 01+→01+ and 01+→02+ decays of 150Nd are evaluated. The effects of particle number projection, static and dynamic deformations, and the full relativistic structure of the transition operators on the matrix elements are studied in detail. Conclusions: The resulting 0 ν β β decay matrix element for the 01+→01+ transition is 5.60 , which gives the most optimistic prediction for the next generation of experiments searching for the 0 ν β β decay in 150Nd.

  10. 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.

  11. 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.

  12. 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.''

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. Quantitative elemental imaging of octopus stylets using PIXE and the nuclear microprobe

    NASA Astrophysics Data System (ADS)

    Doubleday, Zoë; Belton, David; Pecl, Gretta; Semmens, Jayson

    2008-01-01

    By utilising targeted microprobe technology, the analysis of elements incorporated within the hard bio-mineralised structures of marine organisms has provided unique insights into the population biology of many species. As hard structures grow, elements from surrounding waters are incorporated effectively providing a natural 'tag' that is often unique to the animal's particular location or habitat. The spatial distribution of elements within octopus stylets was investigated, using the nuclear microprobe, to assess their potential for determining dispersal and population structure in octopus populations. Proton Induced X-ray Emission (PIXE) was conducted using the Dynamic Analysis method and GeoPIXE software package, which produced high resolution, quantitative elemental maps of whole stylet cross-sections. Ten elements were detected within the stylets which were heterogeneously distributed throughout the microstructure. Although Ca decreased towards the section edge, this trend was consistent between individuals and remained homogeneous in the inner region of the stylet, and thus appears a suitable internal standard for future microprobe analyses. Additional analyses used to investigate the general composition of the stylet structure suggested that they are amorphous and largely organic, however, there was some evidence of phosphatic mineralisation. In conclusion, this study indicates that stylets are suitable for targeted elemental analysis, although this is currently limited to the inner hatch region of the microstructure.

  1. An Overview of the NASA Aviation Safety Program Propulsion Health Monitoring Element

    NASA Technical Reports Server (NTRS)

    Simon, Donald L.

    2000-01-01

    The NASA Aviation Safety Program (AvSP) has been initiated with aggressive goals to reduce the civil aviation accident rate, To meet these goals, several technology investment areas have been identified including a sub-element in propulsion health monitoring (PHM). Specific AvSP PHM objectives are to develop and validate propulsion system health monitoring technologies designed to prevent engine malfunctions from occurring in flight, and to mitigate detrimental effects in the event an in-flight malfunction does occur. A review of available propulsion system safety information was conducted to help prioritize PHM areas to focus on under the AvSP. It is noted that when a propulsion malfunction is involved in an aviation accident or incident, it is often a contributing factor rather than the sole cause for the event. Challenging aspects of the development and implementation of PHM technology such as cost, weight, robustness, and reliability are discussed. Specific technology plans are overviewed including vibration diagnostics, model-based controls and diagnostics, advanced instrumentation, and general aviation propulsion system health monitoring technology. Propulsion system health monitoring, in addition to engine design, inspection, maintenance, and pilot training and awareness, is intrinsic to enhancing aviation propulsion system safety.

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. Heavy-ion double charge exchange reactions: A tool toward 0 νββ nuclear matrix elements

    NASA Astrophysics Data System (ADS)

    Cappuzzello, F.; Cavallaro, M.; Agodi, C.; Bondì, M.; Carbone, D.; Cunsolo, A.; Foti, A.

    2015-11-01

    The knowledge of the nuclear matrix elements for the neutrinoless double beta decay is fundamental for neutrino physics. In this paper, an innovative technique to extract information on the nuclear matrix elements by measuring the cross section of a double charge exchange nuclear reaction is proposed. The basic point is that the initial- and final-state wave functions in the two processes are the same and the transition operators are similar. The double charge exchange cross sections can be factorized in a nuclear structure term containing the matrix elements and a nuclear reaction factor. First pioneering experimental results for the 40Ca(18O,18Ne)40Ar reaction at 270 MeV incident energy show that such cross section factorization reasonably holds for the crucial 0+ → 0+ transition to 40Args, at least at very forward angles.

  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. Epistemic uncertainty in the ranking and categorization of probabilistic safety assessment model elements: issues and findings.

    PubMed

    Borgonovo, Emanuele

    2008-08-01

    In this work, we study the effect of epistemic uncertainty in the ranking and categorization of elements of probabilistic safety assessment (PSA) models. We show that, while in a deterministic setting a PSA element belongs to a given category univocally, in the presence of epistemic uncertainty, a PSA element belongs to a given category only with a certain probability. We propose an approach to estimate these probabilities, showing that their knowledge allows to appreciate "the sensitivity of component categorizations to uncertainties in the parameter values" (U.S. NRC Regulatory Guide 1.174). We investigate the meaning and utilization of an assignment method based on the expected value of importance measures. We discuss the problem of evaluating changes in quality assurance, maintenance activities prioritization, etc. in the presence of epistemic uncertainty. We show that the inclusion of epistemic uncertainly in the evaluation makes it necessary to evaluate changes through their effect on PSA model parameters. We propose a categorization of parameters based on the Fussell-Vesely and differential importance (DIM) measures. In addition, issues in the calculation of the expected value of the joint importance measure are present when evaluating changes affecting groups of components. We illustrate that the problem can be solved using DIM. A numerical application to a case study concludes the work.

  11. 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.

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. 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)

  18. Many-body correlations of QRPA in nuclear matrix elements of double-beta decay

    SciTech Connect

    Terasaki, J.

    2015-10-28

    We present two new ideas on the quasiparticle random-phase approximation (QRPA) approach for calculating nuclear matrix elements of double-beta decay. First, it is necessary to calculate overlaps of the QRPA states obtained on the basis of the ground states of different nuclei. We calculate this overlap using quasiboson vacua as the QRPA ground states. Second, we show that two-particle transfer paths are possible to use for the calculation under the closure approximation. A calculation is shown for {sup 150}Nd→{sup 150}Sm using these two new ideas, and their implication is discussed.

  19. NUCLEAR RADIATION DOSIMETER USING COMPOSITE FILTER AND A SINGLE ELEMENT FILTER

    DOEpatents

    Storm, E.; Shlaer, S.

    1964-04-21

    A nuclear radiation dosimeter is described that uses, in combination, a composite filter and a single element filter. The composite filter contains a plurality of comminuted metals having K-edges evenly distributed over the energy range of interest and the quantity of each of the metals is selected to result in filtering in an amount inversely proportional to the sensitivity of the film in the range over l00 kev. A copper filter is used that has a thickness to contribute the necessary additional correction in the interval between 40 and 100 kev. (AEC)

  20. Numerical Simulations of Single Flow Element in a Nuclear Thermal Thrust Chamber

    NASA Technical Reports Server (NTRS)

    Cheng, Gary; Ito, Yasushi; Ross, Doug; Chen, Yen-Sen; Wang, Ten-See

    2007-01-01

    The objective of this effort is to develop an efficient and accurate computational methodology to predict both detailed and global thermo-fluid environments of a single now element in a hypothetical solid-core nuclear thermal thrust chamber assembly, Several numerical and multi-physics thermo-fluid models, such as chemical reactions, turbulence, conjugate heat transfer, porosity, and power generation, were incorporated into an unstructured-grid, pressure-based computational fluid dynamics solver. The numerical simulations of a single now element provide a detailed thermo-fluid environment for thermal stress estimation and insight for possible occurrence of mid-section corrosion. In addition, detailed conjugate heat transfer simulations were employed to develop the porosity models for efficient pressure drop and thermal load calculations.

  1. Safety of active implantable devices during MRI examinations: a finite element analysis of an implantable pump.

    PubMed

    Büchler, Philippe; Simon, Anne; Burger, Jürgen; Ginggen, Alec; Crivelli, Rocco; Tardy, Yanik; Luechinger, Roger; Olsen, Sigbjørn

    2007-04-01

    The goal of this study was to propose a general numerical analysis methodology to evaluate the magnetic resonance imaging (MRI)-safety of active implants. Numerical models based on the finite element (FE) technique were used to estimate if the normal operation of an active device was altered during MRI imaging. An active implanted pump was chosen to illustrate the method. A set of controlled experiments were proposed and performed to validate the numerical model. The calculated induced voltages in the important electronic components of the device showed dependence with the MRI field strength. For the MRI radiofrequency fields, significant induced voltages of up to 20 V were calculated for a 0.3T field-strength MRI. For the 1.5 and 3.0OT MRIs, the calculated voltages were insignificant. On the other hand, induced voltages up to 11 V were calculated in the critical electronic components for the 3.0T MRI due to the gradient fields. Values obtained in this work reflect to the worst case situation which is virtually impossible to achieve in normal scanning situations. Since the calculated voltages may be removed by appropriate protection circuits, no critical problems affecting the normal operation of the pump were identified. This study showed that the proposed methodology helps the identification of the possible incompatibilities between active implants and MR imaging, and can be used to aid the design of critical electronic systems to ensure MRI-safety.

  2. 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.

  3. Metallographic examination of damaged N reactor spent nuclear fuel element SFEC5,4378

    SciTech Connect

    Marschman, S.C.; Pyecha, T.D.; Abrefah, J.

    1997-08-01

    N-Reactor spent nuclear fuel (SNF) is currently residing underwater in the K Basins at the Hanford site, in Richland, Washington. This report presents results of the metallographic examination of specimens cut from an SNF element (Mark IV-E) with breached cladding. The element had resided in the K-West (KW) Storage Basin for at least 10 years after it was discharged from the N-Reactor. The storage containers in the KW Basin were nominally closed, isolating the SNF elements from the open pool environment. Seven specimens from this Mark IV-E outer fuel element were examined using an optical metallograph. Included were two specimens that had been subjected to a conditioning process recommended by the Independent Technical Assessment Team, two specimens that had been subjected to a conditioning process recommended in the Integrated Process Strategy Report, and three that were in the as-received, as-cut condition. One of the as-received specimens had been cut from the damaged (or breached) end of the element. All other specimens were cut from the undamaged mid-region of the fuel element. The specimens were visually examined to (1) identify uranium hydride inclusions present in the uranium metal fuel, (2) measure the thickness of the oxide layer formed on the uranium edges and assess the apparent integrity and adhesion of the oxide layer, and (3) look for features in the microstructure that might provide an insight into the various corrosion processes that occurred during underwater storage in the KW Basin. These features included, but were not limited to, the integrity of the cladding and the fuel-to-cladding bond, obvious anomalies in the microstructure, excessive pitting or friability of the fuel matrix, and obvious anomalies in the distribution of uranium hydride or uranium carbide inclusions. Also, the observed metallographic features of the conditioned specimens were compared with those of the as-received (unconditioned) specimens. 11 refs., 93 figs., 2 tabs.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Nuclear-Structure Data Relevant to Neutinoless-Double-Beta-Decay Matrix Elements

    NASA Astrophysics Data System (ADS)

    Kay, Benjamin

    2015-10-01

    An observation of neutrinoless double beta decay is one of the most exciting prospects in contemporary physics. It follows that calculations of the nuclear matrix elements for this process are of high priority. The change in the wave functions between the initial and final states of the neutrinoless-double-beta-decay candidates 76Ge-->76Se, 100Mo-->100Ru, 130Te-->130Xe, and 136Xe-->136Ba have been studied with transfer reactions. The data are focused on the change in the occupancies of the valence orbitals in the ground states as two neutrons decay into two protons. The results set a strict constraint on any theoretical calculations describing this rearrangement and thus on the magnitude of the nuclear matrix elements for this process, which currently exhibit uncertainties at the factor of 2-4 level. Prior to these measurements there were limited experimental data were available A = 76 and 100 systems, and very limited data for the A = 130 and 136 systems, in a large part due to the gaseous Xe isotopes involved. The uncertainties on most of these data are estimated to range from 0.1-0.3 nucleons. The program started with the A = 76 system, with subsequent calculations, modified to reproduce the experimental occupancies, exhibiting a significant reduction in the discrepancy between various models. New data are available for the A = 100 , 130, and 136 systems. I review the program, making detailed comparisons between the latest theoretical calculations and the experimental data where available. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Nuclear Physics, under Contract Number DE-AC02-06CH11357.

  12. 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.

  13. Identification of a sequence element directing a protein to nuclear speckles.

    PubMed

    Eilbracht, J; Schmidt-Zachmann, M S

    2001-03-27

    SF3b(155) is an essential spliceosomal protein, highly conserved during evolution. It has been identified as a subunit of splicing factor SF3b, which, together with a second multimeric complex termed SF3a, interacts specifically with the 12S U2 snRNP and converts it into the active 17S form. The protein displays a characteristic intranuclear localization. It is diffusely distributed in the nucleoplasm but highly concentrated in defined intranuclear structures termed "speckles," a subnuclear compartment enriched in small ribonucleoprotein particles and various splicing factors. The primary sequence of SF3b(155) suggests a multidomain structure, different from those of other nuclear speckles components. To identify which part of SF3b(155) determines its specific intranuclear localization, we have constructed expression vectors encoding a series of epitope-tagged SF3b(155) deletion mutants as well as chimeric combinations of SF3b(155) sequences with the soluble cytoplasmic protein pyruvate kinase. Following transfection of cultured mammalian cells, we have identified (i) a functional nuclear localization signal of the monopartite type (KRKRR, amino acids 196--200) and (ii) a molecular segment with multiple threonine-proline repeats (amino acids 208--513), which is essential and sufficient to confer a specific accumulation in nuclear speckles. This latter sequence element, in particular amino acids 208--440, is required for correct subcellular localization of SF3b(155) and is also sufficient to target a reporter protein to nuclear speckles. Moreover, this "speckle-targeting sequence" transfers the capacity for interaction with other U2 snRNP components.

  14. 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.

  15. MOON for neutrino-less {beta}{beta} decays and {beta}{beta} nuclear matrix elements

    SciTech Connect

    Ejiri, H.

    2009-11-09

    The MOON project aims at spectroscopic 0v{beta}{beta} studies with the v-mass sensitivity of 100-30 meV by measuring two beta rays from {sup 100}Mo and/or {sup 82}Se. The detector is a compact super-module of multi-layer PL scintillator plates. R and D works made by the pro to-type MOON-1 and the small PL plate show the possible energy resolution of around {sigma}{approx}2.2%, as required for the mass sensitivity. Nuclear matrix elements M{sup 2v} for 2v{beta}{beta} are shown to be given by the sum {sigma}{sub L}M{sub k} of the 2v{beta}{beta} matrix elements M{sub k} through intermediate quasi-particle states in the Fermi-surface, where Mi is obtained experimentally by using the GT(J{sup {pi}} = 1{sup +}) matrix elements of M{sub i}(k) and M{sub f}(k) for the successive single-{beta} transitions through the k-th intermediate state.

  16. Ionising irradiation alters the dynamics of human long interspersed nuclear elements 1 (LINE1) retrotransposon.

    PubMed

    Tanaka, Atsushi; Nakatani, Youko; Hamada, Nobuyuki; Jinno-Oue, Atsushi; Shimizu, Nobuaki; Wada, Seiichi; Funayama, Tomoo; Mori, Takahisa; Islam, Salequl; Hoque, Sheikh Ariful; Shinagawa, Masahiko; Ohtsuki, Takahiro; Kobayashi, Yasuhiko; Hoshino, Hiroo

    2012-09-01

    It is important to identify the mechanism by which ionising irradiation induces various genomic alterations in the progeny of surviving cells. Ionising irradiation activates mobile elements like retrotransposons, although the mechanism of its phenomena consisting of transcriptions and insertions of the products into new sites of the genome remains unclear. In this study, we analysed the effects of sparsely ionising X-rays and densely ionising carbon-ion beams on the activities of a family of active retrotransposons, long interspersed nuclear elements 1 (L1). We used the L1/reporter knock-in human glioma cell line, NP-2/L1RP-enhanced GFP (EGFP), that harbours full-length L1 tagged with EGFP retrotransposition detection cassette (L1RP-EGFP) in the chromosomal DNA. X-rays and carbon-ion beams similarly increased frequencies the transcription from L1RP-EGFP and its retrotransposition. Short-sized de novo L1RP-EGFP insertions with 5'-truncation were induced by X-rays, while full-length or long-sized insertions (>5 kb, containing ORF1 and ORF2) were found only in cell clones irradiated by the carbon-ion beams. These data suggest that X-rays and carbon-ion beams induce different length of de novo L1 insertions, respectively. Our findings thus highlight the necessity to investigate the mechanisms of mutations caused by transposable elements by ionising irradiation.

  17. MOON for neutrino-less ββ decays and ββ nuclear matrix elements

    NASA Astrophysics Data System (ADS)

    Ejiri, H.

    2009-11-01

    The MOON project aims at spectroscopic 0vββ studies with the v-mass sensitivity of 100-30 meV by measuring two beta rays from 100Mo and/or 82Se. The detector is a compact super-module of multi-layer PL scintillator plates. R&D works made by the pro to-type MOON-1 and the small PL plate show the possible energy resolution of around σ~2.2%, as required for the mass sensitivity. Nuclear matrix elements M2v for 2vββ are shown to be given by the sum ΣLMk of the 2vββ matrix elements Mk through intermediate quasi-particle states in the Fermi-surface, where Mi is obtained experimentally by using the GT(Jπ = 1+) matrix elements of Mi(k) and Mf(k) for the successive single-β transitions through the k-th intermediate state.

  18. 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.

  19. 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.

  20. 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.