National Library of Energy BETA

Sample records for nuclear criticality safety

  1. Autoclave nuclear criticality safety analysis

    SciTech Connect (OSTI)

    D`Aquila, D.M. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Tayloe, R.W. Jr. [Battelle, Columbus, OH (United States)

    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}.

  2. Nuclear criticality safety: 2-day training course

    SciTech Connect (OSTI)

    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.

  3. HANFORD NUCLEAR CRITICALITY SAFETY PROGRAM DATABASE

    SciTech Connect (OSTI)

    TOFFER, H.

    2005-05-02

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

  4. WIPP-025, Rev. 0 Summary of Nuclear Criticality Safety

    E-Print Network [OSTI]

    WIPP-025, Rev. 0 Summary of Nuclear Criticality Safety Evaluation for Shielded Containers AT THE WASTE ISOLATION PILOT PLANT WIPP-025, REV. 0 AUGUST 2009 This document revision was prepared PLANT WIPP-025, REV. 0 AUGUST 2009 Summary of Nuclear Criticality Safety Evaluation for Shielded

  5. Nuclear Criticality Safety Guide for Fire Protection

    Office of Energy Efficiency and Renewable Energy (EERE)

    This guide is intended to provide information for use by fire protection professionals in the application of reasonable methods of fire protection in those facilities where there is a potential for nuclear criticality.

  6. A Web-Based Nuclear Criticality Safety Bibliographic Database

    SciTech Connect (OSTI)

    Koponen, B L; Huang, S

    2007-02-22

    A bibliographic criticality safety database of over 13,000 records is available on the Internet as part of the U.S. Department of Energy's (DOE) Nuclear Criticality Safety Program (NCSP) website. This database is easy to access via the Internet and gets substantial daily usage. This database and other criticality safety resources are available at ncsp.llnl.gov. The web database has evolved from more than thirty years of effort at Lawrence Livermore National Laboratory (LLNL), beginning with compilations of critical experiment reports and American Nuclear Society Transactions.

  7. Nuclear criticality safety engineer qualification program utilizing SAT

    SciTech Connect (OSTI)

    Baltimore, C.J.; Dean, J.C.; Henson, T.L. [Lockheed Martin Utility Services, Inc., Paducah, KY (United States)

    1996-12-31

    As part of the privatization process of the U.S. uranium enrichment plants, the Paducah Gaseous Diffusion Plant (PGDP) and the Portsmouth Gaseous Diffusion Plant (PORTS) have been in transition from U.S. Department of Energy (DOE) regulatory oversight to U.S. Nuclear Regulatory Commission (NRC) oversight since July 1993. One of the focus areas of this transition has been training and qualification of plant personnel who perform tasks important to nuclear safety, such as nuclear criticality safety (NCS) engineers.

  8. Criticality Safety | Department of Energy

    Office of Environmental Management (EM)

    Nuclear Safety Management American Nuclear Society, Nuclear Criticality Safety Division ANSIANS-8 Standards U.S. Department of Energy Nuclear Criticality Safety Program Orders,...

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

    SciTech Connect (OSTI)

    Sanchez, R.G. [comp.

    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.

  10. Nuclear Criticality Safety | More Science | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Safety Program. NCSP is chartered to maintain the technical infrastructure (integral experiments, computational tools, training, data, etc.) needed to support safe,...

  11. WIPP-016, Rev. 0 Nuclear Criticality Safety Evaluation for

    E-Print Network [OSTI]

    WIPP-016, Rev. 0 Nuclear Criticality Safety Evaluation for Contact-Handled Transuranic Waste Department of Energy review required before public release Name/Org: __________/WIPP Project Date 05-HANDLED TRANSURANIC WASTE AT THE WASTE ISOLATION PILOT PLANT WIPP-016, REV. 0 MAY 2008 OFFICIAL USE ONLY Washington

  12. PLC-Based Safety Critical Software Development for Nuclear Power Plants

    E-Print Network [OSTI]

    PLC-Based Safety Critical Software Development for Nuclear Power Plants Junbeom Yoo1 , Sungdeok Cha development technique for nuclear power plants'I&C soft- ware controllers. To improve software safety, we in developing safety-critical control software for a Korean nuclear power plant, and experience to date has been

  13. Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2007-02-07

    This standard provides a framework for generating Criticality Safety Evaluations (CSE) supporting fissionable material operations at Department of Energy (DOE) nonreactor nuclear facilities. This standard imposes no new criticality safety analysis requirements.

  14. Additional nuclear criticality safety calculations for small-diameter containers

    SciTech Connect (OSTI)

    Hone, M.J.

    1996-01-01

    This report documents additional criticality safety analysis calculations for small diameter containers, which were originally documented in Reference 1. The results in Reference 1 indicated that some of the small diameter containers did not meet the criteria established for criticality safety at the Portsmouth facility (K{sub eff} +2{sigma}<.95) when modeled under various contingency assumptions of reflection and moderation. The calculations performed in this report reexamine those cases which did not meet the criticality safety criteria. In some cases, unnecessary conservatism is removed, and in other cases mass or assay limits are established for use with the respective containers.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

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

    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.

  17. Nuclear criticality safety modeling of an LEU deposit

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    LeTellier, M.S.; Smallwood, D.J.; Henkel, J.A. [and others

    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.

  19. Application of Neutron-Absorbing Structural-Amorphous metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls

    E-Print Network [OSTI]

    2006-01-01

    enhance criticality safety for spent nuclear fuel in basketsNuclear Fuel (SNF) Container to Enhance Criticality SafetyNuclear Fuel (SNF) Containers: Use of Novel Coating Materials to Enhance Criticality Safety

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    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 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

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

    SciTech Connect (OSTI)

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

    2014-10-27

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

  4. Applicability of reactor code WIMS for nuclear criticality safety studies

    SciTech Connect (OSTI)

    Matausek, M.V.; Marinkovic, N.

    1995-12-31

    The purpose of this paper is to examine applicability of the reactor code WIMS for calculating criticality parameters of nonreactor configurations containing fissile materials. Results are given and discussed for some typical configurations containing {sup 235}U.

  5. Nuclear criticality safety assessment of the proposed CFC replacement coolants

    SciTech Connect (OSTI)

    Jordan, W.C.; Dyer, H.R.

    1993-12-01

    The neutron multiplication characteristics of refrigerant-114 (R-114) and proposed replacement coolants perfluorobutane (C{sub 4}F{sub 10}) and cycloperfluorobutane C{sub 4}F{sub 8}) have been compared by evaluating the infinite media multiplication factors of UF{sub 6}/H/coolant systems and by replacement calculations considering a 10-MW freezer/sublimer. The results of these comparisons demonstrate that R-114 is a neutron absorber, due to its chlorine content, and that the alternative fluorocarbon coolants are neutron moderators. Estimates of critical spherical geometries considering mixtures of UF{sub 6}/HF/C{sub 4}F{sub 10} indicate that the flourocarbon-moderated systems are large compared with water-moderated systems. The freezer/sublimer calculations indicate that the alternative coolants are more reactive than R-114, but that the reactivity remains significantly below the condition of water in the tubes, which was a limiting condition. Based on these results, the alternative coolants appear to be acceptable; however, several follow-up tasks have been recommended, and additional evaluation will be required on an individual equipment basis.

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

    SciTech Connect (OSTI)

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

    1993-08-16

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

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

    SciTech Connect (OSTI)

    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.

  8. Y-12's 1958 nuclear criticality accident and increased safety - 1958 brought accidents, more safety

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effectWorkingLos Alamos verifies largest single| National1958 nuclear criticality

  9. BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities

    SciTech Connect (OSTI)

    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. Privatization of the gaseous diffusion plants and impacts on nuclear criticality safety administration

    SciTech Connect (OSTI)

    D`Aquila, D.M.; Holliday, R.T. [Lockheed Martin Utility Services, Inc., Piketon, OH (United States); Dean, J.C. [Lockheed Martin Utility Services, Inc., Paducah, KY (United States)

    1996-12-31

    The Energy Policy Act of 1992 created the United States Enrichment Corporation (USEC) on July 1, 1993. The USEC is a government-owned business that leases those Gaseous Diffusion Plant (GDP) facilities at the Portsmouth, Ohio, and Paducah, Kentucky, sites from the U.S. Department of Energy (DOE) that are required for enriching uranium. Lockheed Martin Utility Services is the operating contractor for the USEC-leased facilities. The DOE has retained use of, and regulation over, some facilities and areas at the Portsmouth and Paducah sites for managing legacy wastes and environmental restoration activities. The USEC is regulated by the DOE, but is currently changing to regulation under the U.S. Nuclear Regulatory Commission (NRC). The USEC is also preparing for privatization of the uranium enrichment enterprise. These changes have significantly affected the nuclear criticality safety (NCS) programs at the sites.

  11. Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

    SciTech Connect (OSTI)

    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 keff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

  12. The ORSphere Benchmark Evaluation and Its Potential Impact on Nuclear Criticality Safety

    SciTech Connect (OSTI)

    John D. Bess; Margaret A. Marshall; J. Blair Briggs

    2013-10-01

    In the early 1970’s, critical experiments using an unreflected metal sphere of highly enriched uranium (HEU) were performed with the focus to provide a “very accurate description…as an ideal benchmark for calculational methods and cross-section data files.” Two near-critical configurations of the Oak Ridge Sphere (ORSphere) were evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results from those benchmark experiments were then compared with additional unmoderated and unreflected HEU metal benchmark experiment configurations currently found in the ICSBEP Handbook. For basic geometries (spheres, cylinders, and slabs) the eigenvalues calculated using MCNP5 and ENDF/B-VII.0 were within 3 of their respective benchmark values. There appears to be generally a good agreement between calculated and benchmark values for spherical and slab geometry systems. Cylindrical geometry configurations tended to calculate low, including more complex bare HEU metal systems containing cylinders. The ORSphere experiments do not calculate within their 1s uncertainty and there is a possibility that the effect of the measured uncertainties for the GODIVA I benchmark may need reevaluated. There is significant scatter in the calculations for the highly-correlated ORCEF cylinder experiments, which are constructed from close-fitting HEU discs and annuli. Selection of a nuclear data library can have a larger impact on calculated eigenvalue results than the variation found within calculations of a given experimental series, such as the ORCEF cylinders, using a single nuclear data set.

  13. Nuclear criticality safety calculations for a K-25 site vacuum cleaner

    SciTech Connect (OSTI)

    Shor, J.T.; Haire, M.J.

    1997-02-01

    A modified Nilfisk model GSJ dry vacuum cleaner is used throughout the K-25 Site to collect dry forms of highly enriched uranium (HEU). When vacuuming, solids are collected in a cyclone-type separator vacuum cleaner body. Calculations were done with the SCALE (KENO V.a) computer code to establish conditions at which a nuclear criticality event might occur if the vacuum cleaner was filled with fissile solution. Conditions evaluated included full (12-in. water) reflection and nominal (1-in. water) reflection, and full (100%) and 20% {sup 235}U enrichment. Validation analyses of SCALE/KENO and the SCALE 27-group cross sections for nuclear criticality safety applications indicate that a calculated k{sub eff} + 2{sigma} < 0.9605 may be considered safely subcritical. Thus, a system with a calculated k{sub eff} + 2{sigma} {ge} 0.9605 is considered unsafe and may be critical. Critical conditions were calculated to be 70 g U/L for 100% {sup 235}U and full 12-in. water reflection. This corresponds to a minimum critical mass of approximately 1,400 g {sup 235}U for the approximate 20.0-L volume of the vacuum cleaner. The actual volume of the vacuum cleaner is smaller than the modeled volume because some internal materials of construction were assumed to be fissile solution. The model was an overestimate, for conservatism, of fissile solution occupancy. At nominal reflection conditions, the critical concentration in a vacuum cleaner full of UO{sub 2}F{sub 2} solution was calculated to be 100 g{sup 235}U/L, or 2,000 g mass of 100% {sup 235}U. At 20% {sup 235}U for the 20.0-L volume of the vacuum cleaner. At 15% {sup 235}U enrichment and full reflection, critical conditions were not reached at any possible concentration of uranium as a uranyl fluoride solution. At 17.5% {sup 235}U enrichment, criticality was reached at approximately 1,300 g U/L which is beyond saturation at 25 C.

  14. 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 (OSTI)

    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.

  15. Nuclear Criticality Safety Evaluation of the 9965, 9968, 9972, 9973, 9974, and 9975 Shipping Casks

    SciTech Connect (OSTI)

    Frost, R.L.

    1999-02-26

    A Nuclear Criticality Safety Evaluation (NCSE) has been performed for the 9965, 9968, 9972, 9973, 9974, and 9975 SRS-designed shipping casks. This was done in support of the recertification effort for the 9965 and 9968, and the certification of the newly designed 9972-9975 series. The analysis supports the use of these packages as Fissile Class I for shipment of fissionable material from the SRS FB-Line, HB-Line, and from Lawrence Livermore national Laboratory. six different types of material were analyzed with varying Isotopic composition, of both oxide and metallic form. The mass limits required to support the fissile Class I rating for each of the envelopes are given in the Table below. These mass limits apply if DOE approves an exception as described in 10 CFR 71.55(c), such that water leakage into the primary containment vessel does not need to be considered in the criticality analysis. If this exception is not granted, the mass limits are lower than those shown below. this issue is discussed in detail in sections 5 and 6 of the report.One finding from this work is important enough to highlight in the abstract. The fire tests performed for this family of shipping casks indicates only minimal charring of the Celotex thermal insulation. Analysis of the casks with no Celotex insulation (assuming it has all burned away), results in values of k-eff that exceed 1.0. Therefore, the Celotex insulation must remain intact in order to guarantee sub criticality of the 9972-9975 family of shipping casks.

  16. Nuclear criticality safety controls for uranium deposits during D and D at the Oak Ridge Gaseous Diffusion Plant

    SciTech Connect (OSTI)

    Haire, M.J.; Jordan, W.C. [Oak Ridge National Lab., TN (United States); Jollay, L.J. III; Dahl, T.L. [Oak Ridge K-25 Site, TN (United States)

    1997-02-01

    The US Department of Energy (DOE) Deputy Assistant Secretary of Energy for Environmental Management has issued a challenge to complete DOE environmental cleanup within a decade. The response for Oak Ridge facilities is in accordance with the DOE ten-year plan which calls for completion of > 95% of environmental management work by the year 2006. This will result in a 99% risk reduction and in a significant savings in base line costs in waste management (legacy waste); remedial action (groundwater, soil, etc.); and decontamination and decommissioning (D and D). It is assumed that there will be long-term institutional control of cascade equipment, i.e., there will be no walk away from sites, and that there will be firm radioactivity release limits by 1999 for recycle metals. An integral part of these plants is the removal of uranium deposits which pose nuclear criticality safety concerns in the shut down of the Oak Ridge Gaseous Diffusion Plant. DOE has initiated the Nuclear Criticality Stabilization Program to improve nuclear criticality safety by removing the larger uranium deposits from unfavorable geometry equipment. Nondestructive assay (NDA) measurements have identified the location of these deposits. The objective of the K-25 Site Nuclear Criticality Stabilization Program is to remove and place uranium deposits into safe geometry storage containers to meet the double contingency principle. Each step of the removal process results in safer conditions where multiple controls are present. Upon completion of the Program, nuclear criticality risks will be greatly reduced.

  17. New Resolved Resonance Region Evaluation for 63Cu and 65Cu for Nuclear Criticality Safety Program

    SciTech Connect (OSTI)

    Sobes, Vladimir; Leal, Luiz C; Guber, Klaus H; Forget, Benoit; Kopecky, S.; Schillebeeckx, P.; Siegler, P.

    2014-01-01

    A new resolved resonance region evaluation of 63Cu and 65Cu was done in the energy region from 10-5 eV to 99.5 keV. The R-Matrix SAMMY method using the Reich-Moore approximation was used to create a new set of consistent resonance parameters. The new evaluation was based on three experimental transmission data sets; two measured at ORELA and one from MITR, and two radiative capture experimental data sets from GELINA. A total of 141 new resonances were identied for 63Cu and 117 for 65Cu. The corresponding set of external resonances for each isotope was based on the identied resonances above 99.5 keV from the ORELA transmission data. The negative external levels (bound levels) were determined to match the dierential thermal cross section measured at the MITR. Double dierential elastic scattering cross sections were calculated from the new set of resonance parameters. Benchmarking calculations were carried out on a set of ICSBEP benchmarks. This work is in support of the DOE Nuclear Criticality Safety Program.

  18. Validation of nuclear criticality safety software and 27 energy group ENDF/B-IV cross sections. Revision 1

    SciTech Connect (OSTI)

    Lee, B.L. Jr. [Battelle, Columbus, OH (United States); D`Aquila, D.M. [Lockheed Martin Utility Services, Inc., Oak Ridge, TN (United States)

    1996-01-01

    The original validation report, POEF-T-3636, was documented in August 1994. The document was based on calculations that were executed during June through August 1992. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This revision is written to clarify the margin of safety being used at Portsmouth for nuclear criticality safety calculations. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Lockheed Martin 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 IBM3090 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. For calculations of Portsmouth systems using the specified codes and systems covered by this validation, a maximum k{sub eff} including 2{sigma} of 0.9605 or lower shall be considered as subcritical to ensure a calculational margin of safety of 0.02. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25.

  19. Nuclear Safety | Department of Energy

    Office of Environmental Management (EM)

    Nuclear Safety Nuclear Safety The Office of Nuclear Safety establishes and maintains nuclear safety policy, requirements, and guidance including policy and requirements relating to...

  20. Nuclear Data for Criticality Safety and Reactor Applications at the Gaerttner LINAC Center Y. Danon, R.M. Bahran, E.J. Blain, A.M. Daskalakis, B.J. McDermott, D.G. Williams

    E-Print Network [OSTI]

    Danon, Yaron

    Nuclear Data for Criticality Safety and Reactor Applications at the Gaerttner LINAC Center Y. Danon used in reactor and nuclear criticality safety applications. The goal of this program is to provide to nuclear criticality, neutron shielding applications, nuclear reactor design, and to better understand

  1. 2011 Annual Criticality Safety Program Performance Summary

    SciTech Connect (OSTI)

    Andrea Hoffman

    2011-12-01

    The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection, an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The TSR limits fuel plate bundles to 1085 grams U-235, which is the maximum loading of an ATR fuel element. The overloaded fuel plate bundle contained 1097 grams U-235 and was assembled under an 1100 gram U-235 limit in 1982. In 2003, the limit was reduced to 1085 grams citing a new criticality safety evaluation for ATR fuel elements. The fuel plate bundle inventories were not checked for compliance prior to implementing the reduced limit. A subsequent review of the NMIS inventory did not identify further violations. Requirements Management - The INL Criticality Safety program is organized and well documented. The source requirements for the INL Criticality Safety Program are from 10 CFR 830.204, DOE Order 420.1B, Chapter III, 'Nuclear Criticality Safety,' ANSI/ANS 8-series Industry Standards, and DOE Standards. These source requirements are documented in LRD-18001, 'INL Criticality Safety Program Requirements Manual.' The majority of the criticality safety source requirements are contained in DOE Order 420.1B because it invokes all of the ANSI/ANS 8-Series Standards. DOE Order 420.1B also invokes several DOE Standards, including DOE-STD-3007, 'Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities.' DOE Order 420.1B contains requirements for DOE 'Heads of Field Elements' to approve the criticality safety program and specific elements of the program, namely, the qualification of criticality staff and the method for preparing criticality safety evaluations. This was accomplished by the approval of SAR-400, 'INL Standardized Nuclear Safety Basis Manual,' Chapter 6, 'Prevention of Inadvertent Criticality.' Chapter 6 of SAR-400 contains sufficient detail and/or reference to the specific DOE and contractor documents that adequately describe the INL Criticality Safety Program per the elements specified in DOE Order 420.1B. The Safety Evaluation Report for SAR-400 specifically recognizes that the approval of SAR-400 approves the INL Criticality Safety Program. No new source requirements were released in 2011. A revision to LRD-18001 is

  2. Lecture notes for criticality safety

    SciTech Connect (OSTI)

    Fullwood, R.

    1992-03-01

    These lecture notes for criticality safety are prepared for the training of Department of Energy supervisory, project management, and administrative staff. Technical training and basic mathematics are assumed. The notes are designed for a two-day course, taught by two lecturers. Video tapes may be used at the options of the instructors. The notes provide all the materials that are necessary but outside reading will assist in the fullest understanding. The course begins with a nuclear physics overview. The reader is led from the macroscopic world into the microscopic world of atoms and the elementary particles that constitute atoms. The particles, their masses and sizes and properties associated with radioactive decay and fission are introduced along with Einstein's mass-energy equivalence. Radioactive decay, nuclear reactions, radiation penetration, shielding and health-effects are discussed to understand protection in case of a criticality accident. Fission, the fission products, particles and energy released are presented to appreciate the dangers of criticality. Nuclear cross sections are introduced to understand the effectiveness of slow neutrons to produce fission. Chain reactors are presented as an economy; effective use of the neutrons from fission leads to more fission resulting in a power reactor or a criticality excursion. The six-factor formula is presented for managing the neutron budget. This leads to concepts of material and geometric buckling which are used in simple calculations to assure safety from criticality. Experimental measurements and computer code calculations of criticality are discussed. To emphasize the reality, historical criticality accidents are presented in a table with major ones discussed to provide lessons-learned. Finally, standards, NRC guides and regulations, and DOE orders relating to criticality protection are presented.

  3. Criticality safety basics, a study guide

    SciTech Connect (OSTI)

    V. L. Putman

    1999-09-01

    This document is a self-study and classroom guide, for criticality safety of activities with fissile materials outside nuclear reactors. This guide provides a basic overview of criticality safety and criticality accident prevention methods divided into three parts: theory, application, and history. Except for topic emphasis, theory and history information is general, while application information is specific to the Idaho National Engineering and Environmental Laboratory (INEEL). Information presented here should be useful to personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. However, the guide's primary target audience is fissile material handler candidates.

  4. Office of Nuclear Safety

    Broader source: Energy.gov [DOE]

    The Office of Nuclear Safety establishes nuclear safety requirements and expectations for the Department to ensure protection of workers and the public from the hazards associated with nuclear operations with all Department operations.

  5. Safety-Critical Universit at

    E-Print Network [OSTI]

    Peleska, Jan - Fachbereich 3

    . Hazard Analysis and Risk Assessment 5. Design Criteria for Safety-Critical Systems 6. Validation, Veri#12. Hazard Analysis and Risk Assessment 5. Design Criteria for Safety-Critical Systems 6. Validation, Veri#12Safety-Critical Systems Prof. Dr. Jan Peleska Universit at Bremen | TZI Dr. Ing. Cornelia Zahlten

  6. Nuclear Explosive Safety Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14

    This Manual provides supplemental details to support the requirements of DOE O 452.2D, Nuclear Explosive Safety.

  7. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2014-07-10

    The Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations (NEOs).

  8. Reference handbook: Nuclear criticality

    SciTech Connect (OSTI)

    Not Available

    1991-12-06

    The purpose for this handbook is to provide Rocky Flats personnel with the information necessary to understand the basic principles underlying a nuclear criticality.

  9. Hybrid Statistical Model Checking Technique for Reliable Safety Critical Systems

    E-Print Network [OSTI]

    @cs.kaist.ac.kr Abstract--Reliability of safety critical systems such as nuclear power plants and automobiles has become1 Hybrid Statistical Model Checking Technique for Reliable Safety Critical Systems Youngjoo Kim a significant issue to our society. As more computing systems are utilized in these safety critical systems

  10. Promulgating Nuclear Safety Requirements

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-05-15

    Applies to all Nuclear Safety Requirements Adopted by the Department to Govern the Conduct of its Nuclear Activities. Cancels DOE P 410.1. Canceled by DOE N 251.85.

  11. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14

    This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1D, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations (NEOs). Cancels DOE O 452.2C. Admin Chg 1, dated 7-10-13, cancels DOE O 452.2D.

  12. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2015-01-26

    This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1E, Nuclear Explosive and Weapon Surety Program, or successor directive, for routine and planned nuclear explosive operations (NEOs). Supersedes DOE O 452.2D and DOE M 452.2-1A.

  13. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14

    This Order establishes requirements to implement the nuclear explosive safety elements of DOE O 452.1D, Nuclear Explosive and Weapon Surety Program, for routine and planned nuclear explosive operations. Cancels DOE O 452.2C. Admin Chg 1, 7-10-13

  14. CRITICALITY SAFETY CONTROLS AND THE SAFETY BASIS AT PFP

    SciTech Connect (OSTI)

    Kessler, S

    2009-04-21

    With the implementation of DOE Order 420.1B, Facility Safety, and DOE-STD-3007-2007, 'Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities', a new requirement was imposed that all criticality safety controls be evaluated for inclusion in the facility Documented Safety Analysis (DSA) and that the evaluation process be documented in the site Criticality Safety Program Description Document (CSPDD). At the Hanford site in Washington State the CSPDD, HNF-31695, 'General Description of the FH Criticality Safety Program', requires each facility develop a linking document called a Criticality Control Review (CCR) to document performance of these evaluations. Chapter 5, Appendix 5B of HNF-7098, Criticality Safety Program, provided an example of a format for a CCR that could be used in lieu of each facility developing its own CCR. Since the Plutonium Finishing Plant (PFP) is presently undergoing Deactivation and Decommissioning (D&D), new procedures are being developed for cleanout of equipment and systems that have not been operated in years. Existing Criticality Safety Evaluations (CSE) are revised, or new ones written, to develop the controls required to support D&D activities. Other Hanford facilities, including PFP, had difficulty using the basic CCR out of HNF-7098 when first implemented. Interpretation of the new guidelines indicated that many of the controls needed to be elevated to TSR level controls. Criterion 2 of the standard, requiring that the consequence of a criticality be examined for establishing the classification of a control, was not addressed. Upon in-depth review by PFP Criticality Safety staff, it was not clear that the programmatic interpretation of criterion 8C could be applied at PFP. Therefore, the PFP Criticality Safety staff decided to write their own CCR. The PFP CCR provides additional guidance for the evaluation team to use by clarifying the evaluation criteria in DOE-STD-3007-2007. In reviewing documents used in classifying controls for Nuclear Safety, it was noted that DOE-HDBK-1188, 'Glossary of Environment, Health, and Safety Terms', defines an Administrative Control (AC) in terms that are different than typically used in Criticality Safety. As part of this CCR, a new term, Criticality Administrative Control (CAC) was defined to clarify the difference between an AC used for criticality safety and an AC used for nuclear safety. In Nuclear Safety terms, an AC is a provision relating to organization and management, procedures, recordkeeping, assessment, and reporting necessary to ensure safe operation of a facility. A CAC was defined as an administrative control derived in a criticality safety analysis that is implemented to ensure double contingency. According to criterion 2 of Section IV, 'Linkage to the Documented Safety Analysis', of DOESTD-3007-2007, the consequence of a criticality should be examined for the purposes of classifying the significance of a control or component. HNF-PRO-700, 'Safety Basis Development', provides control selection criteria based on consequence and risk that may be used in the development of a Criticality Safety Evaluation (CSE) to establish the classification of a component as a design feature, as safety class or safety significant, i.e., an Engineered Safety Feature (ESF), or as equipment important to safety; or merely provides defense-in-depth. Similar logic is applied to the CACs. Criterion 8C of DOE-STD-3007-2007, as written, added to the confusion of using the basic CCR from HNF-7098. The PFP CCR attempts to clarify this criterion by revising it to say 'Programmatic commitments or general references to control philosophy (e.g., mass control or spacing control or concentration control as an overall control strategy for the process without specific quantification of individual limits) is included in the PFP DSA'. Table 1 shows the PFP methodology for evaluating CACs. This evaluation process has been in use since February of 2008 and has proven to be simple and effective. Each control identified i

  15. Office of Nuclear Facility Safety Programs

    Broader source: Energy.gov [DOE]

    The Office of Nuclear Facility Safety Programs establishes nuclear safety requirements related to safety management programs that are essential to the safety of DOE nuclear facilities.

  16. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2006-06-12

    The directive provides supplemental details to support the requirements of DOE O 452.2C, Nuclear Explosive Safety, dated 6-12-06. Canceled by DOE M 452.2-1A.

  17. NuDE: Development Environment for Safety-Critical Software of

    E-Print Network [OSTI]

    NuDE: Development Environment for Safety-Critical Software of Nuclear Power Plant Jong-Hoon Lee #12;Overview of NuDE NuDE: Development Environment for Safety-Critical Software of Nuclear Power PlantDE: Development Environment for Safety-Critical Software of Nuclear Power Plant 14 #12;· FTA for Requirements

  18. Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2006-06-12

    The directive establishes specific nuclear explosive safety (NES) program requirements to implement the DOE NES standards and other NES criteria for routine and planned nuclear explosive operations. Cancels DOE O 452.2B. Canceled by DOE O 452.2D.

  19. A nuclear criticality safety assessment of the loss of moderation control in 2 1/2 and 10-ton cylinders containing enriched UF{sub 6}

    SciTech Connect (OSTI)

    Newvahner, R.L. [Martin Marietta Energy Systems, Inc., Piketon, OH (United States); Pryor, W.A. [PAI Corp., Oak Ridge, TN (United States)

    1991-12-31

    Moderation control for maintaining nuclear criticality safety in 2 {1/2}-ton, 10-ton, and 14-ton cylinders containing enriched uranium hexafluoride (UF{sub 6}) has been used safely within the nuclear industry for over thirty years, and is dependent on cylinder integrity and containment. This assessment evaluates the loss of moderation control by the breaching of containment and entry of water into the cylinders. The first objective of this study was to estimate the required amounts of water entering these large UF{sub 6} cylinders to react with, and to moderate the uranium compounds sufficiently to cause criticality. Hypothetical accident situations were modeled as a uranyl fluoride (UO{sub 2}F{sub 2}) slab above a UF{sub 6} hemicylinder, and a UO{sub 2}F{sub 2} sphere centered within a UF{sub 6} hemicylinder. These situations were investigated by computational analyses utilizing the KENO V.a Monte Carlo Computer Code. The results were used to estimate both the masses of water required for criticality, and the limiting masses of water that could be considered safe. The second objective of the assessment was to calculate the time available for emergency control actions before a criticality would occur, i.e., a {open_quotes}safetime{close_quotes}, for various sources of water and different size openings in a breached cylinder. In the situations considered, except the case for a fire hose, the safetime appears adequate for emergency control actions. The assessment shows that current practices for handling moderation controlled cylinders of low enriched UF{sub 6}, along with the continuation of established personnel training programs, ensure nuclear criticality safety for routine and emergency operations.

  20. Nuclear Explosive Safety Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14

    This Department of Energy (DOE) Manual provides supplemental details on selected topics to support the requirements of DOE O 452.2D, Nuclear Explosive Safety, dated 4/14/09. Cancels DOE M 452.2-1. Admin Chg 1, dated 7-10-13, cancels DOE M 452.2-1A.

  1. Nuclear Explosive Safety Evaluation Processes

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14

    This Manual provides supplemental details to support the nuclear explosive safety evaluation requirement of DOE O 452.2D, Nuclear Explosive Safety. Does not cancel other directives. Admin Chg 1, 7-10-13.

  2. Use of a Web Site to Enhance Criticality Safety Training

    SciTech Connect (OSTI)

    Huang, S T; Morman, J

    2003-08-04

    Currently, a website dedicated to enhancing communication and dissemination of criticality safety information is sponsored by the U.S. Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). This website was developed as part of the DOE response to the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2, which reflected the need to make criticality safety information available to a wide audience. The website is the focal point for DOE nuclear criticality safety (NCS) activities, resources and references, including hyperlinks to other sites actively involved in the collection and dissemination of criticality safety information. The website is maintained by the Lawrence Livermore National Laboratory (LLNL) under auspices of the NCSP management. One area of the website contains a series of Nuclear Criticality Safety Engineer Training (NCSET) modules. During the past few years, many users worldwide have accessed the NCSET section of the NCSP website and have downloaded the training modules as an aid for their training programs. This trend was remarkable in that it points out a continuing need of the criticality safety community across the globe. It has long been recognized that training of criticality safety professionals is a continuing process involving both knowledge-based training and experience-based operations floor training. As more of the experienced criticality safety professionals reach retirement age, the opportunities for mentoring programs are reduced. It is essential that some method be provided to assist the training of young criticality safety professionals to replenish this limited human expert resource to support on-going and future nuclear operations. The main objective of this paper is to present the features of the NCSP website, including its mission, contents, and most importantly its use for the dissemination of training modules to the criticality safety community. We will discuss lessons learned and several ideas for future development in the area of web-based training for criticality safety professionals. Our effort is intended to stimulate a discussion of ideas and solicit participation in the development of the next generation of a web-based criticality training site that can be used to assist the training of newcomers to this important safety discipline.

  3. Nuclear explosive safety study process

    SciTech Connect (OSTI)

    NONE

    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.

  4. In-Situ Radiological Surveys to Address Nuclear Criticality Safety Requirements During Remediation Activities at the Shallow Land Disposal Area, Armstrong County, Pennsylvania - 12268

    SciTech Connect (OSTI)

    Norris, Phillip; Mihalo, Mark; Eberlin, John; Lambert, Mike; Matthews, Brian

    2012-07-01

    Cabrera Services Inc. (CABRERA) is the remedial contractor for the Shallow Land Disposal Area (SLDA) Site in Armstrong County Pennsylvania, a United States (US) Army Corps of Engineers - Buffalo District (USACE) contract. The remediation is being completed under the USACE's Formerly Utilized Sites Remedial Action Program (FUSRAP) which was established to identify, investigate, and clean up or control sites previously used by the Atomic Energy Commission (AEC) and its predecessor, the Manhattan Engineer District (MED). As part of the management of the FUSRAP, the USACE is overseeing investigation and remediation of radiological contamination at the SLDA Site in accordance with the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA), 42 US Code (USC), Section 9601 et. seq, as amended and, the National Oil and Hazardous Substance Pollution Contingency Plan (NCP), Title 40 of the Code of Federal Regulations (CFR) Section 300.430(f) (2). The objective of this project is to clean up radioactive waste at SLDA. The radioactive waste contains special nuclear material (SNM), primarily U-235, in 10 burial trenches, Cabrera duties include processing, packaging and transporting the waste to an offsite disposal facility in accordance with the selected remedial alternative as defined in the Final Record of Decision (USACE, 2007). Of particular importance during the remediation is the need to address nuclear criticality safety (NCS) controls for the safe exhumation and management of waste containing fissile materials. The partnership between Cabrera Services, Inc. and Measutronics Corporation led to the development of a valuable survey tool and operating procedure that are essential components of the SLDA Criticality Safety and Material Control and Accountability programs. Using proven existing technologies in the design and manufacture of the Mobile Survey Cart, the continued deployment of the Cart will allow for an efficient and reliable methodology to allow for the safe exhumation of the Special Nuclear Material in existing SLDA trenches. (authors)

  5. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-10-24

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.

  6. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-11-16

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.

  7. Safety Lifecycle for Developing Safety Critical Artificial Neural Networks

    E-Print Network [OSTI]

    Kelly, Tim

    Safety Lifecycle for Developing Safety Critical Artificial Neural Networks Zeshan Kurd, Tim Kelly.kelly}@cs.york.ac.uk Abstract. Artificial neural networks are employed in many areas of industry such as medicine and defence. There are many techniques that aim to improve the performance of neural networks for safety-critical systems

  8. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    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.

  9. DOE Cites Safety and Ecology Corp. for Violating Nuclear Safety...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    DOE Cites Safety and Ecology Corp. for Violating Nuclear Safety Rules DOE Cites Safety and Ecology Corp. for Violating Nuclear Safety Rules June 14, 2005 - 4:53pm Addthis...

  10. CRAD, Facility Safety- Nuclear Facility Safety Basis

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Nuclear Facility Safety Basis.

  11. Safety Activities on Safety-Critical Software for Reactor Protection System Gee-Yong Park1

    E-Print Network [OSTI]

    Jee, Eunkyoung

    Instrumentation & Control Systems) project in order to be used in newly-constructed nuclear power plants and alsoSafety Activities on Safety-Critical Software for Reactor Protection System Gee-Yong Park1 , Kee, 373-1 Guseong, Yuseong, Daejon, 305-701 KOREA INTRODUCTION A fully-digitalized reactor protection

  12. Nuclear Explosive Safety Evaluation Processes

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2009-04-14

    This Manual provides supplemental details to support the nuclear explosive safety (NES) evaluation requirement of Department of Energy (DOE) Order (O) 452.2D, Nuclear Explosive Safety, dated 4/14/09. Admin Chg 1, dated 7-10-13, cancels DOE M 452.2-2.

  13. Nuclear reactor safety device

    DOE Patents [OSTI]

    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. Preservation and Dissemination of the Hardcopy Documentation Portion of the NCSP Nuclear Criticality Bibliographic Database

    SciTech Connect (OSTI)

    Koponen, B L; Heinrichs, D

    2009-05-18

    The U.S. Department of Energy supports a nuclear criticality safety bibliographic internet database that contains approximately 15,000 records. We are working to ensure that a substantial portion of the corresponding hardcopy documents are preserved, digitized, and made available to criticality safety practitioners via the Nuclear Criticality Safety Program web site.

  15. Independent Activity Report, Defense Nuclear Facilities Safety...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Defense Nuclear Facilities Safety Board Public Meeting - October 2012 Independent Activity Report, Defense Nuclear Facilities Safety Board Public Meeting - October 2012 October...

  16. Criticality Safety Basics for INL FMHs and CSOs

    SciTech Connect (OSTI)

    V. L. Putman

    2012-04-01

    Nuclear power is a valuable and efficient energy alternative in our energy-intensive society. However, material that can generate nuclear power has properties that require this material be handled with caution. If improperly handled, a criticality accident could result, which could severely harm workers. This document is a modular self-study guide about Criticality Safety Principles. This guide's purpose it to help you work safely in areas where fissionable nuclear materials may be present, avoiding the severe radiological and programmatic impacts of a criticality accident. It is designed to stress the fundamental physical concepts behind criticality controls and the importance of criticality safety when handling fissionable materials outside nuclear reactors. This study guide was developed for fissionable-material-handler and criticality-safety-officer candidates to use with related web-based course 00INL189, BEA Criticality Safety Principles, and to help prepare for the course exams. These individuals must understand basic information presented here. This guide may also be useful to other Idaho National Laboratory personnel who must know criticality safety basics to perform their assignments safely or to design critically safe equipment or operations. This guide also includes additional information that will not be included in 00INL189 tests. The additional information is in appendices and paragraphs with headings that begin with 'Did you know,' or with, 'Been there Done that'. Fissionable-material-handler and criticality-safety-officer candidates may review additional information at their own discretion. This guide is revised as needed to reflect program changes, user requests, and better information. Issued in 2006, Revision 0 established the basic text and integrated various programs from former contractors. Revision 1 incorporates operation and program changes implemented since 2006. It also incorporates suggestions, clarifications, and additional information from readers and from personnel who took course 00INL189. Revision 1 also completely reorganized the training to better emphasize physical concepts behind the criticality controls that fissionable material handlers and criticality safety officers must understand. The reorganization is based on and consistent with changes made to course 00INL189 due to a review of course exam results and to discussions with personnel who conduct area-specific training.

  17. FAQS Qualification Card – Criticality Safety

    Broader source: Energy.gov [DOE]

    A key element for the Department’s Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA).

  18. CRAD, Criticality Safety- Idaho Accelerated Retrieval Project Phase II

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2006 Commencement of Operations assessment of the Criticality Safety program at the Idaho Accelerated Retrieval Project Phase II.

  19. Criticality Safety Code Validation with LWBR’s SB Cores

    SciTech Connect (OSTI)

    Putman, Valerie Lee

    2003-01-01

    The first set of critical experiments from the Shippingport Light Water Breeder Reactor Program included eight, simple geometry critical cores built with 233UO2-ZrO2, 235UO2-ZrO2, ThO2, and ThO2-233UO2 nuclear materials. These cores are evaluated, described, and modeled to provide benchmarks and validation information for INEEL criticality safety calculation methodology. In addition to consistency with INEEL methodology, benchmark development and nuclear data are consistent with International Criticality Safety Benchmark Evaluation Project methodology.Section 1 of this report introduces the experiments and the reason they are useful for validating some INEEL criticality safety calculations. Section 2 provides detailed experiment descriptions based on currently available experiment reports. Section 3 identifies criticality safety validation requirement sources and summarizes requirements that most affect this report. Section 4 identifies relevant hand calculation and computer code calculation methodologies used in the experiment evaluation, benchmark development, and validation calculations. Section 5 provides a detailed experiment evaluation. This section identifies resolutions for currently unavailable and discrepant information. Section 5 also reports calculated experiment uncertainty effects. Section 6 describes the developed benchmarks. Section 6 includes calculated sensitivities to various benchmark features and parameters. Section 7 summarizes validation results. Appendices describe various assumptions and their bases, list experimenter calculations results for items that were independently calculated for this validation work, report other information gathered and developed by SCIENTEC personnel while evaluating these same experiments, and list benchmark sample input and miscellaneous supplementary data.

  20. Nuclear Reactor Safety Design Criteria

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1993-01-19

    The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Supersedes DOE 5480.1, dated 1-19-93. Certified 11-18-10.

  1. An assessment of criticality safety at the Department of Energy Rocky Flats Plant, Golden, Colorado, July--September 1989

    SciTech Connect (OSTI)

    Mattson, Roger J.

    1989-09-01

    This is a report on the 1989 independent Criticality Safety Assessment of the Rocky Flats Plant, primarily in response to public concerns that nuclear criticality accidents involving plutonium may have occurred at this nuclear weapon component fabrication and processing plant. The report evaluates environmental issues, fissile material storage practices, ventilation system problem areas, and criticality safety practices. While no evidence of a criticality accident was found, several recommendations are made for criticality safety improvements. 9 tabs.

  2. Use of InSpector{sup TM} 1 1000 Instrument with LaBr{sub 3} for Nuclear Criticality Safety (NCS) Applications at the Westinghouse Hematite Decommissioning Project (HDP) - 13132

    SciTech Connect (OSTI)

    Pritchard, Megan [Nuclear Safety Associates, P.O. Box 471488, Charlotte, NC 28247 (United States)] [Nuclear Safety Associates, P.O. Box 471488, Charlotte, NC 28247 (United States); Guido, Joe [System One Services, 12 Federal St. Ste. 205, Pittsburgh, PA 15212 (United States)] [System One Services, 12 Federal St. Ste. 205, Pittsburgh, PA 15212 (United States)

    2013-07-01

    The Westinghouse Hematite Decommissioning Project (HDP) is a former nuclear fuel cycle facility that is currently undergoing decommissioning. One aspect of the decommissioning scope is remediation of buried nuclear waste in unlined burial pits. The current Nuclear Criticality Safety program relies on application of criticality controls based on radiological setpoints from a 2 x 2 Sodium Iodide (NaI) detector. Because of the nature of the material buried (Low Enriched Uranium (LEU), depleted uranium, thorium, and radium) and the stringent threshold for application of criticality controls based on waste management (0.1 g {sup 235}U/L), a better method for {sup 235}U identification and quantification has been developed. This paper outlines the early stages of a quick, in-field nuclear material assay and {sup 235}U mass estimation process currently being deployed at HDP. Nuclear material initially classified such that NCS controls are necessary can be demonstrated not to require such controls and dispositioned as desired by project operations. Using Monte Carlo techniques and a high resolution Lanthanum Bromide (LaBr) detector with portable Multi-Channel Analyzer (MCA), a bounding {sup 235}U mass is assigned to basic geometries of nuclear material as it is excavated. The deployment of these methods and techniques has saved large amounts of time and money in the nuclear material remediation process. (authors)

  3. NRC - regulator of nuclear safety

    SciTech Connect (OSTI)

    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.

  4. Safety culture assessment based on PSA-defined critical components

    SciTech Connect (OSTI)

    Mavko, B.; Kozuh, M.

    1994-12-31

    With the suggested guide-words approach connected to the critical components, a different viewpoint on nuclear safety attitudes is defined. This enables the identification, judgment, and improvement of the most vulnerable places in the plant. Any potential overlap in the duties and areas where a clear division of responsibilities is needed is thus revealed. Also, the need for communication between different groups becomes evident. It is known that anyone who neglects the communication of component status by assuming everybody knows it can cause a serious problem. Safety culture is reached when such assumptions are absent from day-to-day operations.

  5. Safety of Nuclear Explosive Operations

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2001-08-07

    This directive establishes responsibilities and requirements to ensure the safety of routine and planned nuclear explosive operations and associated activities and facilities. Cancels DOE O 452.2A and DOE G 452.2A-1A. Canceled by DOE O 452.2C.

  6. Nuclear Safety Information Agreement Between the U.S. Nuclear...

    Broader source: Energy.gov (indexed) [DOE]

    Environment, Health, Safety and Security (EHSS DOE), Cathy Haney (Director, Office of Nuclear Materials Safety and Safeguards (NRC)), Marissa Bailey (Director, Division of Fuel...

  7. WIPP Documents - Nuclear Safety

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDidDevelopmentat LENA|UpcomingVisit UsNewsNational EnvironmentalNuclear

  8. Criticality Safety Basics for INL Emergency Responders

    SciTech Connect (OSTI)

    Valerie L. Putman

    2012-08-01

    This document is a modular self-study guide about criticality safety principles for Idaho National Laboratory emergency responders. This guide provides basic criticality safety information for people who, in response to an emergency, might enter an area that contains much fissionable (or fissile) material. The information should help responders understand unique factors that might be important in responding to a criticality accident or in preventing a criticality accident while responding to a different emergency.

    This study guide specifically supplements web-based training for firefighters (0INL1226) and includes information for other Idaho National Laboratory first responders. However, the guide audience also includes other first responders such as radiological control personnel.

    For interested readers, this guide includes clearly marked additional information that will not be included on tests. The additional information includes historical examples (Been there. Done that.), as well as facts and more in-depth information (Did you know …).

    INL criticality safety personnel revise this guide as needed to reflect program changes, user requests, and better information. Revision 0, issued May 2007, established the basic text. Revision 1 incorporates operation, program, and training changes implemented since 2007. Revision 1 increases focus on first responders because later responders are more likely to have more assistance and guidance from facility personnel and subject matter experts. Revision 1 also completely reorganized the training to better emphasize physical concepts behind the criticality controls that help keep emergency responders safe. The changes are based on and consistent with changes made to course 0INL1226.

  9. Nuclear Safety | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuelsof EnergyApril 2014Department ofWind CareerEnergy NuclearNuclear Safety

  10. CRAD, Nuclear Safety Delegations for Documented Safety Analysis...

    Broader source: Energy.gov (indexed) [DOE]

    January 8, 2015 Nuclear Safety Delegations for Documented Safety Analysis Approval (EA CRAD 31-09, Rev. 0) This Criteria Review and Approach Document (EA CRAD 31-09, Rev. 0)...

  11. Spent Nuclear Fuel Project Safety Management Plan

    SciTech Connect (OSTI)

    Garvin, L.J.

    1996-02-01

    The Spent Nuclear Fuel Project Safety Management Plan describes the new nuclear facility regulatory requirements basis for the Spemt Nuclear Fuel (SNF) Project and establishes the plan to achieve compliance with this basis at the new SNF Project facilities.

  12. Validation of Criticality Safety Calculations with SCALE 6.2

    SciTech Connect (OSTI)

    Marshall, William BJ J [ORNL] [ORNL; Wiarda, Dorothea [ORNL] [ORNL; Celik, Cihangir [ORNL] [ORNL; Rearden, Bradley T [ORNL] [ORNL

    2013-01-01

    SCALE 6.2 provides numerous updates in nuclear data, nuclear data processing, and computational tools utilized in the criticality safety calculational sequences relative to SCALE 6.1. A new 252-group ENDF/B-VII.0 multigroup neutron library, improved ENDF/B-VII.0 continuous energy data, as well as the previously deployed 238-group ENDF/B-VII.0 neutron library are included in SCALE 6.2 for criticality safety analysis. The performance of all three libraries for keff calculations is examined with a broad sampling of critical experiment models covering a range of fuels and moderators. Critical experiments from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) that are available in the SCALE Verified, Archived Library of Inputs and Data (VALID) are used in this validation effort. Over 300 cases are used in the validation of KENO V.a, and a more limited set of approximately 50 configurations are used for KENO-VI validation. Additionally, some KENO V.a cases are converted to KENO-VI models so that an equivalent set of experiments can be used to validate both codes. For continuous-energy calculations, SCALE 6.2 provides improved performance relative to SCALE 6.1 in most areas with notable improvements in fuel pin lattice cases, particularly those with mixed oxide fuel. Multigroup calculations with the 252-group library also demonstrate improved performance for fuel lattices, uranium (high and intermediate enrichment) and plutonium metal experiments, and plutonium solution systems. Overall, SCALE 6.2 provides equivalent or smaller biases than SCALE 6.1, and the two versions of KENO provide similar results on the same suite of problems.

  13. CRAD, NNSA- Nuclear Explosive Safety (NES)

    Broader source: Energy.gov [DOE]

    CRAD for Nuclear Explosive Safety (NES). Criteria Review and Approach Documents (CRADs) that can be used to conduct a well-organized and thorough assessment of elements of safety and health programs.

  14. Identification of Integral Benchmarks for Nuclear Data Testing Using DICE (Database for the International Handbook of Evaluated Criticality Safety Benchmark Experiments)

    SciTech Connect (OSTI)

    J. Blair Briggs; A. Nichole Ellis; Yolanda Rugama; Nicolas Soppera; Manuel Bossant

    2011-08-01

    Typical users of the International Criticality Safety Evaluation Project (ICSBEP) Handbook have specific criteria to which they desire to find matching experiments. Depending on the application, those criteria may consist of any combination of physical or chemical characteristics and/or various neutronic parameters. The ICSBEP Handbook contains a structured format helping the user narrow the search for experiments of interest. However, with nearly 4300 different experimental configurations and the ever increasing addition of experimental data, the necessity to perform multiple criteria searches have rendered these features insufficient. As a result, a relational database was created with information extracted from the ICSBEP Handbook. A users’ interface was designed by OECD and DOE to allow the interrogation of this database. The database and the corresponding users’ interface are referred to as DICE. DICE currently offers the capability to perform multiple criteria searches that go beyond simple fuel, physical form and spectra and includes expanded general information, fuel form, moderator/coolant, neutron-absorbing material, cladding, reflector, separator, geometry, benchmark results, spectra, and neutron balance parameters. DICE also includes the capability to display graphical representations of neutron spectra, detailed neutron balance, sensitivity coefficients for capture, fission, elastic scattering, inelastic scattering, nu-bar and mu-bar, as well as several other features.

  15. A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems

    E-Print Network [OSTI]

    . INTRODUCTION Safety-critical systems (e.g. nuclear power plants and air- planes) require rigorous quality a domain-specific point of view. In the RPS (Reactor Protection System) in nuclear power plants, the mostA Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon

  16. Criticality safety evaluation for Portsmouth X-345 High-Enriched-Uranium storage area

    SciTech Connect (OSTI)

    Koponen, B.L.

    1993-09-20

    This report evaluates nuclear criticality safety for the High-Enriched Uranium storage area of the X-345 building of the Portsmouth Gaseous Diffusion Plant. The effects of loss of moderation or mass control are examined for storage units in or out of the storage receptacles. Recommendations are made for decreasing criticality hazards under some conditions of storage or handling considered to be hazardous.

  17. Formal Methods for the Specification and Design of RealTime Safety Critical

    E-Print Network [OSTI]

    Ostroff, Jonathan S.

    and control safety critical sys­ tems. Real­time software controls aircraft, shuts down nuclear power reac of computers in such systems offers considerable benefits, but also poses serious risks to life­time constraints: if the temperature of a nuclear reactor core is too high an alarm must be generated within some

  18. Criticality safety aspects of K-25 Building uranium deposit removal

    SciTech Connect (OSTI)

    Haire, M.J.; Jordan, W.C. [Oak Ridge National Lab., TN (United States); Ingram, J.C. III; Stinnet, E.C. Jr. [Oak Ridge K-25 Site, TN (United States)

    1995-12-31

    The K-25 Building of the Oak Ridge Gaseous Diffusion Plant (now the K-25 Site) went into operation during World War II as the first large scale production plant to separate {sup 235}U from uranium by the gaseous diffusion process. It operated successfully until 1964, when it was placed in a stand-by mode. The Department of Energy has initiated a decontamination and decommissioning program. The primary objective of the Deposit Removal (DR) Project is to improve the nuclear criticality safety of the K-25 Building by removing enriched uranium deposits from unfavorable-geometry process equipment to below minimum critical mass. The method utilized to accomplish this are detailed in this report.

  19. Nuclear Safety Information Agreement Between the U.S. Nuclear...

    Office of Environmental Management (EM)

    for the two agencies to exchange information related to safety issues associated with non-reactor nuclear facilities that would be beneficial to both agencies in performance of...

  20. The history of nuclear weapon safety devices

    SciTech Connect (OSTI)

    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.

  1. Criticality Safety Controls Implementation, May 31, 2013 (HSS...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    representative lines of inquiry to assess criticality control implementation as an integral part of the review of the core functions and implementation of integrated safety...

  2. CRAD, Nuclear Safety Component - June 29, 2011 | Department of...

    Office of Environmental Management (EM)

    Nuclear Safety Component - June 29, 2011 CRAD, Nuclear Safety Component - June 29, 2011 June 29, 2011 Nuclear Safety Component and Services Procurement (HSS CRAD 45-12, Rev. 1)...

  3. RECENT ADDITIONS OF CRITICALITY SAFETY RELATED INTEGRAL BENCHMARK DATA TO THE ICSBEP AND IRPHEP HANDBOOKS

    SciTech Connect (OSTI)

    J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Sartori

    2009-09-01

    High-quality integral benchmark experiments have always been a priority for criticality safety. However, interest in integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of future criticality safety needs to support next generation reactor and advanced fuel cycle concepts. The importance of drawing upon existing benchmark data is becoming more apparent because of dwindling availability of critical facilities worldwide and the high cost of performing new experiments. Integral benchmark data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the International Handbook of Reactor Physics Benchmark Experiments are widely used. Benchmark data have been added to these two handbooks since the last Nuclear Criticality Safety Division Topical Meeting in Knoxville, Tennessee (September 2005). This paper highlights these additions.

  4. A Desktop 3D Printer in Safety-Critical Java

    E-Print Network [OSTI]

    A Desktop 3D Printer in Safety-Critical Java Tórur Biskopstø Strøm Kongens Lyngby 2012 IMM-MSc-2012-critical use cases implemented according to the specification. This thesis presents a RepRap 3D desktop printer

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

    SciTech Connect (OSTI)

    Marshall, A.C.; Lee, J.H.; McCulloch, W.H. (Sandia National Labs., Albuquerque, NM (United States)); Sawyer, J.C. Jr. (National Aeronautics and Space Administration, Washington, DC (United States)); Bari, R.A. (Brookhaven National Lab., Upton, NY (United States)); Brown, N.W. (General Electric Co., San Jose, CA (United States)); 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. NSPWG-recommended safety requirements and guidelines for SEI nuclear propulsion

    SciTech Connect (OSTI)

    Marshall, A.C.; Lee, J.H.; McCulloch, W.H. [Sandia National Labs., Albuquerque, NM (United States); Sawyer, J.C. Jr. [National Aeronautics and Space Administration, Washington, DC (United States); Bari, R.A. [Brookhaven National Lab., Upton, NY (United States); Brown, N.W. [General Electric Co., San Jose, CA (United States); Cullingford, H.S.; Hardy, A.C. [National Aeronautics and Space Administration, Houston, TX (United States). Lyndon B. Johnson Space Center; Niederauer, G.F. [Los Alamos National Lab., NM (United States); Remp, K. [National Aeronautics and Space Administration, Cleveland, OH (United States). Lewis Research Center; Rice, J.W. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Sholtis, J.A. [Department of the Air Force, Kirtland AFB, NM (United States)

    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.

  7. Neutron absorbing coating for nuclear criticality control

    DOE Patents [OSTI]

    Mizia, Ronald E. (Idaho Falls, ID); Wright, Richard N. (Idaho Falls, ID); Swank, William D. (Idaho Falls, ID); Lister, Tedd E. (Idaho Falls, ID); Pinhero, Patrick J. (Idaho Falls, ID)

    2007-10-23

    A neutron absorbing coating for use on a substrate, and which provides nuclear criticality control is described and which includes a nickel, chromium, molybdenum, and gadolinium alloy having less than about 5% boron, by weight.

  8. Nuclear Safety Research and Development Status Workshop Summary

    Office of Environmental Management (EM)

    Chief of Nuclear Safety Staff NUCLEAR SAFETY R&D Perform a peer review of Risk Assessment Corporation WTP analysis by a team and identify Using other benchmarked...

  9. Nuclear Safety Research and Development Annual Report, December...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Nuclear Safety Research and Development Annual Report, December 2014 Nuclear Safety Research and Development Annual Report, December 2014 December 8, 2014 This document is the...

  10. Defense Nuclear Facilities Safety Board (DNFSB) Update - Dale...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Defense Nuclear Facilities Safety Board (DNFSB) Update - Dale Govan, Departmental Representative to the DNFSB Defense Nuclear Facilities Safety Board (DNFSB) Update - Dale Govan,...

  11. CRAD, New Nuclear Facility Documented Safety Analysis and Technical...

    Broader source: Energy.gov (indexed) [DOE]

    December 2, 2014 New Nuclear Facility Documented Safety Analysis and Technical Safety Requirements Criteria Review and Approach Document (EA CRAD 31-07, Rev. 0) CRAD, New Nuclear...

  12. Criticality Safety Evaluation of Hanford Tank Farms Facility

    SciTech Connect (OSTI)

    WEISS, E.V.

    2000-12-15

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste.

  13. New Criticality Safety Analysis Capabilities in SCALE 5.1

    SciTech Connect (OSTI)

    Bowman, Stephen M [ORNL; DeHart, Mark D [ORNL; Dunn, Michael E [ORNL; Goluoglu, Sedat [ORNL; Horwedel, James E [ORNL; Petrie Jr, Lester M [ORNL; Rearden, Bradley T [ORNL; Williams, Mark L [ORNL

    2007-01-01

    Version 5.1 of the SCALE computer software system developed at Oak Ridge National Laboratory, released in 2006, contains several significant enhancements for nuclear criticality safety analysis. This paper highlights new capabilities in SCALE 5.1, including improved resonance self-shielding capabilities; ENDF/B-VI.7 cross-section and covariance data libraries; HTML output for KENO V.a; analytical calculations of KENO-VI volumes with GeeWiz/KENO3D; new CENTRMST/PMCST modules for processing ENDF/B-VI data in TSUNAMI; SCALE Generalized Geometry Package in NEWT; KENO Monte Carlo depletion in TRITON; and plotting of cross-section and covariance data in Javapeno.

  14. Nuclear Safety at the Department of Energy

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2011-12-05

    Nuclear Safety is a core value of the Department of Energy. As our management principle state: "We will pursue our mission in a manner that is safe, secure, legally and ethically sound, and fiscally responsible."

  15. 2012 Nuclear Safety Workshop | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Background In response to the March 2011 accident at the Fukushima Daiichi nuclear power plant, Secretary Chu initiated a series of actions to review the safety of the Department...

  16. Criticality Safety Validation of SCALE 6.1 with ENDF/B-VII.0 Libraries

    SciTech Connect (OSTI)

    Marshall, William BJ J [ORNL] [ORNL; Rearden, Bradley T [ORNL] [ORNL

    2012-01-01

    ANSI/ANS-8.1-1998;2007, Nuclear Criticality Safety in Operations with Fissionable Material Outside Reactors, and ANSI/ANS-8.24-2007, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, require validation of a computer code and the associated data through benchmark evaluations based on physical experiments. The performance of the code and data are validated by comparing the calculated and the benchmark results. A SCALE procedure has been established to generate a Verified, Archived Library of Inputs and Data (VALID). This procedure provides a framework for preparing, peer reviewing, and controlling models and data sets derived from benchmark definitions so that the models and data can be used with confidence. The procedure ensures that the models and data were correctly generated using appropriate references with documented checks and reviews. Configuration management is implemented to prevent inadvertent modification of the models and data or inclusion of models that have not been subjected to the rigorous review process. VALID entries for criticality safety are based on critical experiments documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). The findings of a criticality safety validation of SCALE 6.1 utilizing the benchmark models vetted in the VALID library at Oak Ridge National Laboratory are summarized here.

  17. Nuclear Explosive Safety Study Functional Area Qualification Standard

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2010-05-27

    A Nuclear Explosive Safety Study (NESS) is performed on all DOE Nuclear Explosive Operations (NEOs) in accordance with DOE O 452.1D, Nuclear Explosive and Weapon Surety Program; DOE O 452.2D, Nuclear Explosive Safety; and DOE M 452.2-2, Nuclear Explosive Safety Evaluation Processes.

  18. A Critical Step Toward Sustainable Nuclear Fuel Disposal | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    that the United States remains at the forefront of nuclear safety and security, non-proliferation, and nuclear energy technology we must develop an effective strategy and...

  19. Office of Nuclear Safety Basis and Facility Design

    Broader source: Energy.gov [DOE]

    The Office of Nuclear Safety Basis & Facility Design establishes safety basis and facility design requirements and expectations related to analysis and design of nuclear facilities to ensure protection of workers and the public from the hazards associated with nuclear operations.

  20. CRAD, Criticality Safety- Los Alamos National Laboratory TA 55 SST Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for an assessment of the Criticality Safety program at the Los Alamos National Laboratory, TA 55 SST Facility.

  1. CRAD, Criticality Safety- Y-12 Enriched Uranium Operations Oxide Conversion Facility

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a January 2005 assessment of the Criticality Safety program at the Y-12 - Enriched Uranium Facility.

  2. Fault tree synthesis for software design analysis of PLC based safety-critical systems

    SciTech Connect (OSTI)

    Koo, S. R.; Cho, C. H. [Corporate R and D Inst., Doosan Heavy Industries and Construction Co., Ltd., 39-3, Seongbok-Dong, Yongin-Si, Gyeonggi-Do 449-795 (Korea, Republic of); Seong, P. H. [Dept. of Nuclear and Quantum Engineering, Korea Advanced Inst. of Science and Technology, 373-3 Guseong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

    2006-07-01

    As a software verification and validation should be performed for the development of PLC based safety-critical systems, a software safety analysis is also considered in line with entire software life cycle. In this paper, we propose a technique of software safety analysis in the design phase. Among various software hazard analysis techniques, fault tree analysis is most widely used for the safety analysis of nuclear power plant systems. Fault tree analysis also has the most intuitive notation and makes both qualitative and quantitative analyses possible. To analyze the design phase more effectively, we propose a technique of fault tree synthesis, along with a universal fault tree template for the architecture modules of nuclear software. Consequently, we can analyze the safety of software on the basis of fault tree synthesis. (authors)

  3. Determination of Critical Exponents in Nuclear Systems

    E-Print Network [OSTI]

    W. F. J. Mueller; ALADIN collaboration

    1996-07-08

    Signatures of critical behaviour in nuclear fragmentation are often based on arguments from percolation theory. We demonstrate with general thermodynamic considerations and studies of the Ising model that the reliance on percolation as a reference model bears the risk of missing parts of the essential physics.

  4. Joint nuclear safety research projects between the US and Russian Federation International Nuclear Safety Centers

    SciTech Connect (OSTI)

    Bougaenko, S.E.; Kraev, A.E. [International Nuclear Safety Center of the Russian MINATOM, Moscow (Russian Federation); Hill, D.L.; Braun, J.C.; Klickman, A.E. [Argonne National Lab., IL (United States). International Nuclear Safety Center

    1998-08-01

    The Russian Federation Ministry for Atomic Energy (MINATOM) and the US Department of Energy (USDOE) formed international Nuclear Safety Centers in October 1995 and July 1996, respectively, to collaborate on nuclear safety research. Since January 1997, the two centers have initiated the following nine joint research projects: (1) INSC web servers and databases; (2) Material properties measurement and assessment; (3) Coupled codes: Neutronic, thermal-hydraulic, mechanical and other; (4) Severe accident management for Soviet-designed reactors; (5) Transient management and advanced control; (6) Survey of relevant nuclear safety research facilities in the Russian Federation; (8) Advanced structural analysis; and (9) Development of a nuclear safety research and development plan for MINATOM. The joint projects were selected on the basis of recommendations from two groups of experts convened by NEA and from evaluations of safety impact, cost, and deployment potential. The paper summarizes the projects, including the long-term goals, the implementing strategy and some recent accomplishments for each project.

  5. DOE's Approach to Nuclear Facility Safety Analysis and Management

    Broader source: Energy.gov [DOE]

    Presenter: Dr. James O'Brien, Director, Office of Nuclear Safety, Office of Health, Safety and Security, US Department of Energy

  6. Nuclear Safety Basis Program Review Overview and Management Oversight...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Safety Basis Program Review Overview and Management Oversight Standard Review Plan Nuclear Safety Basis Program Review Overview and Management Oversight Standard Review Plan This...

  7. Nuclear Power - Operation, Safety and Environment 

    E-Print Network [OSTI]

    2011-01-01

    for Advanced Reactors 47 P. F. Frutuoso e Melo, I. M. S. Oliveira and P. L. Saldanha Chapter 4 Geodetic Terrestrial Observations for the Determination of the Stability in the Kr?ko Nuclear Power Plant Region 71 S. Sav?ek, T. Ambro?i? and D. Kogoj Chapter... Experience in Nuclear Steam Reheat 3 Eugene Saltanov and Igor Pioro Chapter 2 Integrated Approach for Actual Safety Analysis 29 Francesco D?Auria, Walter Giannotti and Marco Cherubini Chapter 3 LWR Safety Analysis and Licensing and Implications...

  8. SHEBA-II as a criticality safety benchmark experiment

    SciTech Connect (OSTI)

    LaBauve, R.J.; Sapir, J.L.

    1995-07-01

    SHEBA-II (Solution High Energy Burst Assembly-II) is a critical assembly experiment currently (1995) being operated at the Los Alamos Critical Experiments Facility. It is a bare assembly fueled with an aqueous solution of about 5% enriched uranyl fluoride that is stored in four critically safe steel tanks. The solution is transferred to the critical assembly vessel (CAV) by applying gas pressure to the storage tanks. Reactivity is controlled by varying the solution level, and a safety rod may be inserted in a thimble along the central axis of the CAV for fast shutdown. The simple geometry provided by this cylindrical system allows for easily applied calculational methods, and thus SHEBA-II is ideally suited for use as a criticality safety benchmark experiment.

  9. Management of National Nuclear Power Programs for assured safety

    SciTech Connect (OSTI)

    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).

  10. Sandia Energy - Sandia Nuclear Power Safety Expert Elected to...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Nuclear Power Safety Expert Elected to National Academy of Engineering Home Infrastructure Security Energy Nuclear Energy Capabilities News News & Events Research & Capabilities...

  11. NNSA Cites Los Alamos National Laboratory For Nuclear Safety...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Cites Los Alamos National Laboratory For Nuclear Safety Violations | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the...

  12. Nuclear safety information sharing agreement between NRC and...

    Office of Environmental Management (EM)

    for DOE and NRC to exchange information related to safety issues associated with non-reactor nuclear facilities. The NRC-DOE Inter-Agency nuclear safety information sharing...

  13. Critical Nuclear Charges for N-Electron Atoms

    E-Print Network [OSTI]

    Kais, Sabre

    Critical Nuclear Charges for N-Electron Atoms ALEXEI V. SERGEEV, SABRE KAIS Department of Chemistry is proposed to describe the motion of a loosely bound electron in a multielectron atom when the nuclear charge, which is treated as a continuous parameter, approaches its critical value. The critical nuclear charge

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

    SciTech Connect (OSTI)

    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 joint projects are given.

  15. Safety of Decommissioning of Nuclear Facilities

    SciTech Connect (OSTI)

    Batandjieva, B.; Warnecke, E.; Coates, R. [International Atomic Energy Agency, Vienna (Austria)

    2008-01-15

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

  16. Guide to verification and validation of the SCALE-4 criticality safety software

    SciTech Connect (OSTI)

    Emmett, M.B.; Jordan, W.C.

    1996-12-01

    Whenever a decision is made to newly install the SCALE nuclear criticality safety software on a computer system, the user should run a set of verification and validation (V&V) test cases to demonstrate that the software is properly installed and functioning correctly. This report is intended to serve as a guide for this V&V in that it specifies test cases to run and gives expected results. The report describes the V&V that has been performed for the nuclear criticality safety software in a version of SCALE-4. The verification problems specified by the code developers have been run, and the results compare favorably with those in the SCALE 4.2 baseline. The results reported in this document are from the SCALE 4.2P version which was run on an IBM RS/6000 workstation. These results verify that the SCALE-4 nuclear criticality safety software has been correctly installed and is functioning properly. A validation has been performed for KENO V.a utilizing the CSAS25 criticality sequence and the SCALE 27-group cross-section library for {sup 233}U, {sup 235}U, and {sup 239}Pu fissile, systems in a broad range of geometries and fissile fuel forms. The experimental models used for the validation were taken from three previous validations of KENO V.a. A statistical analysis of the calculated results was used to determine the average calculational bias and a subcritical k{sub eff} criteria for each class of systems validated. Included the statistical analysis is a means of estimating the margin of subcriticality in k{sub eff}. This validation demonstrates that KENO V.a and the 27-group library may be used for nuclear criticality safety computations provided the system being analyzed falls within the range of the experiments used in the validation.

  17. Central Technical Authority Responsibilities Regarding Nuclear Safety Requirements

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2007-08-28

    The order establishes Central Technical Authority and Chief of Nuclear Safety/Chief of Defense Nuclear Safety responsibilities and requirements directed by the Secretary of Energy in the development and issuance of Department of Energy regulations and directives that affect nuclear safety. Does not cancel/supersede other directives.

  18. Reactor and Nuclear Systems Division (RNSD)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    RNSD Home Research Groups Advanced Reactor Systems & Safety Nuclear Data & Criticality Safety Nuclear Security Modeling Radiation Safety Information Computational Center Radiation...

  19. Nonreactor Nuclear Safety Design Criteria and Explosive Safety Criteria Guide for Use with DOE O 420.1, Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-03-28

    This Guide provides guidance on the application of requirements for nonreactor nuclear facilities and explosives facilities of Department of Energy (DOE) O 420.1, Facility Safety, Section 4.1, Nuclear and Explosives Safety Design Criteria. No cancellation.

  20. NEW - DOE O 452.2E, Nuclear Explosive Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of DOE O 452.1D, Nuclear Explosive and Weapon Surety Program, or successor directive, for routine and planned nuclear explosive operations (NEOs).

  1. GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

    SciTech Connect (OSTI)

    J. Blair Briggs; John D. Bess; Jim Gulliford

    2011-09-01

    Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the ICSBEP and the IRPhEP will be discussed in the full paper, selected benchmarks that have been added to the ICSBEP Handbook will be highlighted, and a preview of the new benchmarks that will appear in the September 2011 edition of the Handbook will be provided. Accomplishments of the IRPhEP will also be highlighted and the future of both projects will be discussed. REFERENCES (1) International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03/I-IX, Organisation for Economic Co-operation and Development-Nuclear Energy Agency (OECD-NEA), September 2010 Edition, ISBN 978-92-64-99140-8. (2) International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/DOC(2006)1, Organisation for Economic Co-operation and Development-Nuclear Energy Agency (OECD-NEA), March 2011 Edition, ISBN 978-92-64-99141-5.

  2. Double-clad nuclear fuel safety rod

    DOE Patents [OSTI]

    McCarthy, William H. (Los Altos, CA); Atcheson, Donald B. (Cupertino, CA); Vaidyanathan, Swaminathan (San Jose, CA)

    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.

  3. Nuclear Safety Enforcement Documents | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURINGEnergy Bills and ReduceNovember 2014 Postings(NSUF) GatewayNuclear Safety

  4. Nuclear Safety Regulatory Framework | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURINGEnergy Bills and ReduceNovember 2014 Postings(NSUF) GatewayNuclear SafetyRegulatory

  5. Approach for Validating Actinide and Fission Product Compositions for Burnup Credit Criticality Safety Analyses

    SciTech Connect (OSTI)

    Radulescu, Georgeta; Gauld, Ian C; Ilas, Germina; Wagner, John C

    2014-01-01

    This paper describes a depletion code validation approach for criticality safety analysis using burnup credit for actinide and fission product nuclides in spent nuclear fuel (SNF) compositions. The technical basis for determining the uncertainties in the calculated nuclide concentrations is comparison of calculations to available measurements obtained from destructive radiochemical assay of SNF samples. Probability distributions developed for the uncertainties in the calculated nuclide concentrations were applied to the SNF compositions of a criticality safety analysis model by the use of a Monte Carlo uncertainty sampling method to determine bias and bias uncertainty in effective neutron multiplication factor. Application of the Monte Carlo uncertainty sampling approach is demonstrated for representative criticality safety analysis models of pressurized water reactor spent fuel pool storage racks and transportation packages using burnup-dependent nuclide concentrations calculated with SCALE 6.1 and the ENDF/B-VII nuclear data. The validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

  6. Safety Reports Series No. 11, Developing Safety Culture in Nuclear Activities: Practical Suggestions to Assist Progress, International Atomic Energy Agency

    Office of Energy Efficiency and Renewable Energy (EERE)

    Safety Reports Series No. 11, Developing Safety Culture in Nuclear Activities: Practical Suggestions to Assist Progress, International Atomic Energy Agency

  7. Experience With the SCALE Criticality Safety Cross Section Libraries

    SciTech Connect (OSTI)

    Bowman, S.M.

    2000-08-21

    This report provides detailed information on the SCALE criticality safety cross-section libraries. Areas covered include the origins of the libraries, the data on which they are based, how they were generated, past experience and validations, and performance comparisons with measured critical experiments and numerical benchmarks. The performance of the SCALE criticality safety cross-section libraries on various types of fissile systems are examined in detail. Most of the performance areas are demonstrated by examining the performance of the libraries vs critical experiments to show general trends and weaknesses. In areas where directly applicable critical experiments do not exist, performance is examined based on the general knowledge of the strengths and weaknesses of the cross sections. In this case, the experience in the use of the cross sections and comparisons with the results of other libraries on the same systems are relied on for establishing acceptability of application of a particular SCALE library to a particular fissile system. This report should aid in establishing when a SCALE cross-section library would be expected to perform acceptably and where there are known or suspected deficiencies that would cause the calculations to be less reliable. To determine the acceptability of a library for a particular application, the calculational bias of the library should be established by directly applicable critical experiments.

  8. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-11-20

    The objective of this Order is to establish facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. The Order has Change 1 dated 11-16-95, Change 2 dated 10-24-96, and the latest Change 3 dated 11-22-00 incorporated. The latest change satisfies a commitment made to the Defense Nuclear Facilities Safety Board (DNFSB) in response to DNFSB recommendation 97-2, Criticality Safety.

  9. Recent Use of Covariance Data for Criticality Safety Assessment

    SciTech Connect (OSTI)

    Rearden, Bradley T [ORNL; Mueller, Don [ORNL

    2008-01-01

    The TSUNAMI codes of the Oak Ridge National Laboratory SCALE code system were applied to a burnup credit application to demonstrate the use of sensitivity and uncertainty analysis with recent cross section covariance data for criticality safety code and data validation. The use of sensitivity and uncertainty analysis provides for the assessment of a defensible computational bias, bias uncertainty, and gap analysis for a complex system that otherwise could be assessed only through the use of expert judgment and conservative assumptions.

  10. Critical success factors in implementing process safety management

    SciTech Connect (OSTI)

    Wilson, D.J. [Chevron USA, Inc., New Orleans, LA (United States)

    1996-08-01

    This paper focuses on several {open_quotes}Critical Success Factors {close_quotes} which will determine how well employees will embrace and utilize the changes being asked of them to implement Process Safety Management (PSM). These success factors are applicable to any change which involves asking employees to perform activities differently than they are currently performing them. This includes changes in work processes (the way we arrange and conduct a set of tasks) or changes in work activities (how we perform individual tasks). Simply developing new work processes and explaining them to employees is not enough to ensure that employees will actually utilize them -- no matter how good these processes are. To ensure successful, complete implementation of Process Safety Management, we must manage the transition from how we perform our work now to how we will perform it after PSM is implemented. Environmental and safety performance improvements, facility reliability and operability increases, and employee effectiveness and productivity gains CAN NOT be achieved until Process Safety Management processes are fully implemented. To successfully implement management of change, mechanical integrity, or any of the other processes in PSM, each of the following critical success factors must be carefully considered and utilized as appropriate. They are: (1) Vision of a Future State, Current State Assessment, and a Detailed Plan to Achieve the Future State, (2) Management Commitment, (3) Ownership by Key Individuals, (4) Justification for Actions, (5) Autonomy to Customize the Process, (6) Feedback Mechanism to Adjust Activities, and (7) Process to Refocus & Redirect Efforts.

  11. Review of Nevada Site Office Criticality Safety Assessments at the Criticality Experiments Facility and Training Assembly for Criticality Safety and Appraisal of the Criticality Experiments Facility Startup Plan, October 2011

    Broader source: Energy.gov [DOE]

    This report provides the results of an independent oversight review of criticality safety assessment activities conducted by the Department of Energy's (DOE) Nevada Site Office

  12. Sandia Energy - Sandia Teaches Nuclear Safety Course

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home RoomPreservation of Fe(II)Geothermal EnergyRenewableCompanies PilotTeaches Nuclear Safety Course

  13. Bibliography for nuclear criticality accident experience, alarm systems, and emergency management

    SciTech Connect (OSTI)

    Putman, V.L.

    1995-09-01

    The characteristics, detection, and emergency management of nuclear criticality accidents outside reactors has been an important component of criticality safety for as long as the need for this specialized safety discipline has been recognized. The general interest and importance of such topics receives special emphasis because of the potentially lethal, albeit highly localized, effects of criticality accidents and because of heightened public and regulatory concerns for any undesirable event in nuclear and radiological fields. This bibliography lists references which are potentially applicable to or interesting for criticality alarm, detection, and warning systems; criticality accident emergency management; and their associated programs. The lists are annotated to assist bibliography users in identifying applicable: industry and regulatory guidance and requirements, with historical development information and comments; criticality accident characteristics, consequences, experiences, and responses; hazard-, risk-, or safety-analysis criteria; CAS design and qualification criteria; CAS calibration, maintenance, repair, and testing criteria; experiences of CAS designers and maintainers; criticality accident emergency management (planning, preparedness, response, and recovery) requirements and guidance; criticality accident emergency management experience, plans, and techniques; methods and tools for analysis; and additional bibliographies.

  14. Application of the SCALE TSUNAMI Tools for the Validation of Criticality Safety Calculations Involving 233U

    SciTech Connect (OSTI)

    Mueller, Don [ORNL; Rearden, Bradley T [ORNL; Hollenbach, Daniel F [ORNL

    2009-02-01

    The Radiochemical Development Facility at Oak Ridge National Laboratory has been storing solid materials containing 233U for decades. Preparations are under way to process these materials into a form that is inherently safe from a nuclear criticality safety perspective. This will be accomplished by down-blending the {sup 233}U materials with depleted or natural uranium. At the request of the U.S. Department of Energy, a study has been performed using the SCALE sensitivity and uncertainty analysis tools to demonstrate how these tools could be used to validate nuclear criticality safety calculations of selected process and storage configurations. ISOTEK nuclear criticality safety staff provided four models that are representative of the criticality safety calculations for which validation will be needed. The SCALE TSUNAMI-1D and TSUNAMI-3D sequences were used to generate energy-dependent k{sub eff} sensitivity profiles for each nuclide and reaction present in the four safety analysis models, also referred to as the applications, and in a large set of critical experiments. The SCALE TSUNAMI-IP module was used together with the sensitivity profiles and the cross-section uncertainty data contained in the SCALE covariance data files to propagate the cross-section uncertainties ({Delta}{sigma}/{sigma}) to k{sub eff} uncertainties ({Delta}k/k) for each application model. The SCALE TSUNAMI-IP module was also used to evaluate the similarity of each of the 672 critical experiments with each application. Results of the uncertainty analysis and similarity assessment are presented in this report. A total of 142 experiments were judged to be similar to application 1, and 68 experiments were judged to be similar to application 2. None of the 672 experiments were judged to be adequately similar to applications 3 and 4. Discussion of the uncertainty analysis and similarity assessment is provided for each of the four applications. Example upper subcritical limits (USLs) were generated for application 1 based on trending of the energy of average lethargy of neutrons causing fission, trending of the TSUNAMI similarity parameters, and use of data adjustment techniques.

  15. Safety Series No. 75-INSAG-4, Safety Culture: A report by the International Nuclear Safety Advisory Group, International Atomic Energy Agency

    Broader source: Energy.gov [DOE]

    Safety Series No. 75-INSAG-4, Safety Culture: A report by the International Nuclear Safety Advisory Group, International Atomic Energy Agency, IAEA, 1991

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

    SciTech Connect (OSTI)

    Liu, Xudong; Lemaitre-Xavier, E.; Ahn, Joonhong [Department of Nuclear Engineering, University of California, Berkeley, Berkeley, CA 94720 (United States); Hirano, Fumio [Japan Atomic Energy Agency, Geological Isolation Research and Development Directorate, Tokai-mura, Ibaraki 319-1194 (Japan)

    2013-07-01

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

  17. Nuclear Safety Component and Services Procurement, June 29, 2011...

    Office of Environmental Management (EM)

    Office of Enforcement and Oversight Criteria Review and Approach Document Subject: Nuclear Safety Component and Services Procurement Inspection Criteria, Inspection Activities, and...

  18. Spent Nuclear Fuel project integrated safety management plan

    SciTech Connect (OSTI)

    Daschke, K.D.

    1996-09-17

    This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

  19. DOE's Nuclear Weapons Complex: Challenges to Safety, Security...

    Broader source: Energy.gov (indexed) [DOE]

    Committee on Energy and Commerce U.S. House of Representatives "DOE's Nuclear Weapons Complex: Challenges to Safety, Security, and Taxpayer Stewardship" FOR RELEASE ON DELIVERY...

  20. Requirements Discovery During the Testing of Safety-Critical Software Robyn R. Lutz

    E-Print Network [OSTI]

    Lutz, Robyn R.

    Requirements Discovery During the Testing of Safety-Critical Software Robyn R. Lutz Jet Propulsion discovery during the testing of a safety-critical software system. Analysis of problem reports generated by the integration and system testing of an embedded, safety-critical software system identified four common

  1. Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Documents

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2014-12-19

    This Standard describes a framework and the criteria to be used for approval of (1) safety basis documents, as required by 10 Code of Federal Regulation (C.F.R.) 830, Nuclear Safety Management, and (2) safety design basis documents, as required by Department of Energy (DOE) Standard (STD)-1189-2008, Integration of Safety into the Design Process.

  2. AUDIT REPORT Follow-up on Nuclear Safety: Safety Basis and Quality...

    Broader source: Energy.gov (indexed) [DOE]

    Nuclear Safety: Safety Basis and Quality Assurance at the Los Alamos National Laboratory DOEIG-0941 July 2015 U.S. Department of Energy Office of Inspector General Office of...

  3. Criticality Safety Controls Implementation Inspection Criteria, Approach, and Lines of Inquiry, October 23, 2009, (HSS CRAD 64-18, Rev 0)

    Office of Energy Efficiency and Renewable Energy (EERE)

    DOE has set expectations for implementing criticality safety controls that are selected to provide preventive and/or mitigative functions for specific potential accident scenarios. There are additional expectations for criticality safety controls that are also designated as Specific Administrative Controls (SACs) (see HSS CRAD 64-32). Also, in instances when the review addresses functionality and operability of structures, systems, and components (SSCs) of nuclear facilities specifically required for criticality safety per the facility's documented safety analysis (DSA), see HSS CRAD 64-11.

  4. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    the entire area of nuclear safety. A portion of the safetypeaceful uses of nuclear energy; health and safety measuresU. S. Nuclear Regulatory Conunission, "Reactor Safety Study:

  5. Organized by Academy for Global Nuclear Safety and Security Agent,

    E-Print Network [OSTI]

    Furui, Sadaoki

    Research Collaboration Center, Tokyo Tech. Fukushima Daiichi Nuclear Power Plants on March 20, 2011Organized by Academy for Global Nuclear Safety and Security Agent, Tokyo Institute of Technology Energy Agency (JAEA) Institute of Nuclear Materials Management (INMM) Japan Chapter International Nuclear

  6. NUCLEAR POWER PLANT Nuclear power plants have safety and security procedures in place and

    E-Print Network [OSTI]

    NUCLEAR POWER PLANT ACCIDENTS Nuclear power plants have safety and security procedures in place and are closely monitored by the Nuclear Regulatory Commission (NRC). An accident at a nuclear power plant could of nuclear power plant accidents? Radioactive materials in the plume from the nuclear power plant can settle

  7. Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program

    SciTech Connect (OSTI)

    Smith, Paul Herrick; Teague, Jonathan Gayle

    2015-04-30

    Criticality safety at TA-55 relies on nuclear material containers that are water resistant to prevent significant amounts of water from coming into contact with fissile material in the event of a fire that causes a breach of glovevbox confinement and subsequent fire water ingress. A “water tight container” is a container that will not allow more than 50ml of water ingress when fully submerged, except when under sufficient pressure to produce structural discontinuity. There are many types of containers, welded containers, hermetically sealed containers, filtered containers, etc.

  8. Accurate fission data for nuclear safety

    E-Print Network [OSTI]

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

    2013-04-08

    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 Jyvaskyla. 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 (10^12 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.

  9. Accurate fission data for nuclear safety

    E-Print Network [OSTI]

    Solders, A; Jokinen, A; Kolhinen, V S; Lantz, M; Mattera, A; Penttila, H; Pomp, S; Rakopoulos, V; Rinta-Antila, S

    2013-01-01

    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 Jyvaskyla. 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 (10^12 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...

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

    SciTech Connect (OSTI)

    Timothy J. Leahy

    2010-06-01

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

  11. 2015 Nuclear & Facility Safety Programs Workshop | Department...

    Office of Environmental Management (EM)

    featuring tracks for the Facility Representative (FR), Safety System Oversight (SSO), Fire Safety (FS) and Readiness communities. Distinguished speakers from inside and outside...

  12. Challenges in the Modeling and Quantitative Analysis of Safety-Critical Automotive Systems!

    E-Print Network [OSTI]

    Leue, Stefan

    Challenges in the Modeling and Quantitative Analysis of Safety-Critical Automotive Systems! #12;Joint work with! 2! Matthias Kuntz and Bernd Reh TRW Automotive Stefan Leue Univeristy of Konstanz #12;Outline! ! Safety-Critical Automotive Systems! Safety Analysis Methods in the Automotive Domain

  13. Review of Nuclear Safety Culture at the Hanford Site Waste Treatment...

    Office of Environmental Management (EM)

    Review of Nuclear Safety Culture at the Hanford Site Waste Treatment and Immobilization Plant Project, October 2010 Review of Nuclear Safety Culture at the Hanford Site Waste...

  14. Activities of the PNC Nuclear Safety Working Group

    SciTech Connect (OSTI)

    Kato, W.Y.

    1991-12-31

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

  15. Activities of the PNC Nuclear Safety Working Group

    SciTech Connect (OSTI)

    Kato, W.Y.

    1991-01-01

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

  16. Consideration of nuclear criticality when disposing of transuranic waste at the Waste Isolation Pilot Plant

    SciTech Connect (OSTI)

    RECHARD,ROBERT P.; SANCHEZ,LAWRENCE C.; STOCKMAN,CHRISTINE T.; TRELLUE,HOLLY R.

    2000-04-01

    Based on general arguments presented in this report, nuclear criticality was eliminated from performance assessment calculations for the Waste Isolation Pilot Plant (WIPP), a repository for waste contaminated with transuranic (TRU) radioisotopes, located in southeastern New Mexico. At the WIPP, the probability of criticality within the repository is low because mechanisms to concentrate the fissile radioisotopes dispersed throughout the waste are absent. In addition, following an inadvertent human intrusion into the repository (an event that must be considered because of safety regulations), the probability of nuclear criticality away from the repository is low because (1) the amount of fissile mass transported over 10,000 yr is predicted to be small, (2) often there are insufficient spaces in the advective pore space (e.g., macroscopic fractures) to provide sufficient thickness for precipitation of fissile material, and (3) there is no credible mechanism to counteract the natural tendency of the material to disperse during transport and instead concentrate fissile material in a small enough volume for it to form a critical concentration. Furthermore, before a criticality would have the potential to affect human health after closure of the repository--assuming that a criticality could occur--it would have to either (1) degrade the ability of the disposal system to contain nuclear waste or (2) produce significantly more radioisotopes than originally present. Neither of these situations can occur at the WIPP; thus, the consequences of a criticality are also low.

  17. A critical review of world jet transport safety

    E-Print Network [OSTI]

    Achtmann, Eric D.

    1995-01-01

    This thesis is intended to serve as a comprehensive introduction to world jet transport safety and aviation fire safety. Divided into six sections, this thesis contains: 1) a statistical review of overall levels of safety ...

  18. Interface with the Defense Nuclear Facilities Safety Board

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-12-30

    The manual defines the process DOE will use to interface with the Defense Nuclear Facilities Safety Board and its staff. Canceled by DOE M 140.1-1A. Does not cancel other directives.

  19. Interface with the Defense Nuclear Facilities Safety Board

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2001-03-30

    This Manual presents the process the Department of Energy will use to interface with the Defense Nuclear Facilities Safety Board (DNFSB) and its staff. Supersedes DOE M 140.1-1A.

  20. Interface with the Defense Nuclear Facilities Safety Board

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-01-26

    This Manual presents the process the Department of Energy will use to interface with the Defense Nuclear Facilities Safety Board (DNFSB) and its staff. Cancels DOE M 140.1-1.

  1. Preparation Of Nonreactor Nuclear Facility Documented Safety Analysis

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2014-11-12

    This Department of Energy (DOE) Standard (STD), DOE-STD-3009-2014, describes a method for preparing a Documented Safety Analysis (DSA) that is acceptable to DOE for nonreactor nuclear facilities.

  2. Safety Oversight of Decommissioning Activities at DOE Nuclear Sites

    SciTech Connect (OSTI)

    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.

  3. Nuclear Safety Design Principles & the Concept of Independence: Insights from Nuclear Weapon Safety for Other High-Consequence Applications.

    SciTech Connect (OSTI)

    Brewer, Jeffrey D.

    2014-05-01

    Insights developed within the U.S. nuclear weapon system safety community may benefit system safety design, assessment, and management activities in other high consequence domains. The approach of assured nuclear weapon safety has been developed that uses the Nuclear Safety Design Principles (NSDPs) of incompatibility, isolation, and inoperability to design safety features, organized into subsystems such that each subsystem contributes to safe system responses in independent and predictable ways given a wide range of environmental contexts. The central aim of the approach is to provide a robust technical basis for asserting that a system can meet quantitative safety requirements in the widest context of possible adverse or accident environments, while using the most concise arrangement of safety design features and the fewest number of specific adverse or accident environment assumptions. Rigor in understanding and applying the concept of independence is crucial for the success of the approach. This paper provides a basic description of the assured nuclear weapon safety approach, in a manner that illustrates potential application to other domains. There is also a strong emphasis on describing the process for developing a defensible technical basis for the independence assertions between integrated safety subsystems.

  4. Preliminary Criticality Safety Evaluation for In Situ Grouting in the Subsurface Disposal Area

    SciTech Connect (OSTI)

    Slate, Lawrence J; Taylor, Joseph Todd

    2000-08-01

    A preliminary criticality safety evaluation is presented for in situ grouting in the Subsurface Disposal Area (SDA) at the Idaho National Engineering Laboratory. The grouting materials evaluated are cement and paraffin. The evaluation determines physical and administrative controls necessary to preclude criticality and identifies additional information required for a final criticality safety evaluation. The evaluation shows that there are no criticality concerns with cementitious grout but a neutron poison such as boron would be required for the use of the paraffin matrix.

  5. Nonreactor Nuclear Safety Design Guide for use with DOE O 420.1C, Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2012-12-04

    This Guide provides an acceptable approach for safety design of DOE hazard category 1, 2 and 3 nuclear facilities for satisfying the requirements of DOE O 420.1C. Supersedes DOE G 420.1-1.

  6. Seminar in Critical Inquiry Twenty-first Century Nuclear Systems

    SciTech Connect (OSTI)

    LeMone, D. V.

    2002-02-25

    Critical Inquiry, has not only been successful in increasing university student retention rate but also in improving student academic performance beyond the initial year of transition into the University. The seminar course herein reviewed is a balanced combination of student personal and academic skill development combined with a solid background in modern nuclear systems. It is a valid premise to assume that entering students as well as stakeholders of the general public demonstrate equal levels of capability. Nuclear systems is designed to give a broad and basic knowledge of nuclear power, medical, industrial, research, and military systems (nuclear systems) in 20-25 hours.

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

    SciTech Connect (OSTI)

    Morris, Tommy J. [Los Alamos National Laboratory

    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.

  8. Nuclear safety for the space exploration initiative. Final report

    SciTech Connect (OSTI)

    Dix, T.E.

    1991-11-01

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

  9. A Desktop 3D Printer in Safety-Critical Java Trur Biskopst Strm

    E-Print Network [OSTI]

    Schoeberl, Martin

    A Desktop 3D Printer in Safety-Critical Java Tórur Biskopstø Strøm Department of Informatics according to the specification. In this paper we present a 3D printer and its safety-critical Java level 1 evaluate the specification by implementing a RepRap 3D desktop printer as a use case. A RepRap is a desktop

  10. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2005-12-22

    The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and the System Engineer Program.Chg 1 incorporates the use of DOE-STD-1189-2008, Integration of Safety into the Design Process, mandatory for Hazard Category 1, 2 and 3 nuclear facilities. Cancels DOE O 420.1A.

  11. FAQS Qualification Card – Nuclear Explosive Safety Study

    Broader source: Energy.gov [DOE]

    A key element for the Department’s Technical Qualification Programs is a set of common Functional Area Qualification Standards (FAQS) and associated Job Task Analyses (JTA). These standards are developed for various functional areas of responsibility in the Department, including oversight of safety management programs identified as hazard controls in Documented Safety Analyses (DSA).

  12. NEW - DOE O 420.1 Chg 1, Facility Safety - DOE Directives, Delegations...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena...

  13. Critical Nuclear Utilities Upgrade Project (CNUUP) (4572)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submit theCovalent Bonding in Actinide SandwichCray era Craycourse-inventoryCritical

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

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    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 andmore »activities are controlled through regulatory requirements and supported by Probabilistic Safety Assessment (PSA).« less

  15. CRITICALITY SAFETY OF PROCESSING SALT SOLUTION AT SRS

    SciTech Connect (OSTI)

    Stephens, K; Davoud Eghbali, D; Michelle Abney, M

    2008-01-15

    High level radioactive liquid waste generated as a result of the production of nuclear material for the United States defense program at the Savannah River Site has been stored as 36 million gallons in underground tanks. About ten percent of the waste volume is sludge, composed of insoluble metal hydroxides primarily hydroxides of Mn, Fe, Al, Hg, and most radionuclides including fission products. The remaining ninety percent of the waste volume is saltcake, composed of primarily sodium (nitrites, nitrates, and aluminates) and hydroxides. Saltcakes account for 30% of the radioactivity while the sludge accounts for 70% of the radioactivity. A pilot plant salt disposition processing system has been designed at the Savannah River Site for interim processing of salt solution and is composed of two facilities: the Actinide Removal Process Facility (ARPF) and the Modular Caustic Side Solvent Extraction Unit (MCU). Data from the pilot plant salt processing system will be used for future processing salt at a much higher rate in a new salt processing facility. Saltcake contains significant amounts of actinides, and other long-lived radioactive nuclides such as strontium and cesium that must be extracted prior to disposal as low level waste. The extracted radioactive nuclides will be mixed with the sludge from waste tanks and vitrified in another facility. Because of the presence of highly enriched uranium in the saltcake, there is a criticality concern associated with concentration and/or accumulation of fissionable material in the ARP and MCU.

  16. Exploratory Nuclear Reactor Safety Analysis and Visualization...

    Office of Scientific and Technical Information (OSTI)

    via Integrated Topological and Geometric Techniques A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming...

  17. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Charges Relating to Nuclear Reactor Safety," 1976, availablestudies of light-water nuclear reactor safety, emphasizingstudies of overall nuclear reactor safety have been

  18. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    discussed, between the Nuclear safety assurance and riskCharges Relating to Nuclear Reactor Safety," 1976, availableof light-water nuclear reactor safety, emphasizing the

  19. CRAD, Facility Safety- Nuclear Facility Design

    Office of Energy Efficiency and Renewable Energy (EERE)

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Nuclear Facility Design.

  20. Reducing the Cost of Redundant Execution in Safety-Critical Systems using Relaxed Dedication

    E-Print Network [OSTI]

    Huang, Wei

    Reducing the Cost of Redundant Execution in Safety-Critical Systems using Relaxed Dedication Brett to execute non-critical tasks when critical tasks are not executing. Relaxed dedication is one such technique variety of applications and architectures, relaxed dedication is more cost-effective than a traditional

  1. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-10-13

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. Cancels DOE 5480.7A, DOE 5480.24, DOE 5480.28 and Division 13 of DOE 6430.1A. Canceled by DOE O 420.1A.

  2. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2005-12-22

    This Order establishes facility and programmatic safety requirements for Department of Energy facilities, which includes nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards mitigation, and the System Engineer Program. Cancels DOE O 420.1A. DOE O 420.1B Chg 1 issued 4-19-10.

  3. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2012-12-04

    The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. Cancels DOE O 420.1B, DOE G 420.1-2 and DOE G 420.1-3.

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

    SciTech Connect (OSTI)

    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.

  5. Department of Energy Nuclear Safety Policy

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2011-02-08

    It is the policy of the Department of Energy to design, construct, operate, and decommission its nuclear facilities in a manner that ensures adequate protection of workers, the public, and the environment. Supersedes SEN-35-91.

  6. Optimization of Lyapunov invariants in analysis and implementation of safety-critical software systems

    E-Print Network [OSTI]

    Roozbehani, Mardavij

    2008-01-01

    This dissertation contributes to two major research areas in safety-critical software systems, namely, software analysis, and software implementation. In reference to the software analysis problem, the main contribution ...

  7. Critical processes and performance measures for patient safety systems in healthcare institutions: a Delphi study 

    E-Print Network [OSTI]

    Akins, Ralitsa B.

    2004-11-15

    This dissertation study presents a conceptual framework for implementing and assessing patient safety systems in healthcare institutions. The conceptual framework consists of critical processes and performance measures identified in the context...

  8. An Integrated Formal Approach for Developing High Quality Software for Safety-Critical Systems

    E-Print Network [OSTI]

    Ouyang, Meng

    This report presents the results of a study which devises an Integrated Formal Approach (IFA) for improving specifications of the designs of computer programs used in safety-critical systems. In this IFA, the formal ...

  9. Reducing nuclear danger through intergovernmental technical exchanges on nuclear materials safety management

    SciTech Connect (OSTI)

    Jardine, L.J. [Lawrence Livermore National Lab., CA (United States); Peddicord, K.L. [Texas A and M Univ., College Station, TX (United States); Witmer, F.E.; Krumpe, P.F. [USDOE, Washington, DC (United States); Lazarev, L.; Moshkov, M. [Radievyj Inst., Leningrad (Russian Federation)

    1997-04-09

    The United States and Russia are dismantling nuclear weapons and generating hundreds of tons of excess plutonium and high enriched uranium fissile nuclear materials that require disposition. The U.S. Department of Energy and Russian Minatom organizations.are planning and implementing safe, secure storage and disposition operations for these materials in numerous facilities. This provides a new opportunity for technical exchanges between Russian and Western scientists that can establish an improved and sustained common safety culture for handling these materials. An initiative that develops and uses personal relationships and joint projects among Russian and Western participants involved in fissile nuclear materials safety management contributes to improving nuclear materials nonproliferation and to making a safer world. Technical exchanges and workshops are being used to systematically identify opportunities in the nuclear fissile materials facilities to improve and ensure the safety of workers, the public, and the environment.

  10. Safety management of nuclear waste in Spain

    SciTech Connect (OSTI)

    Echavarri, L.E. (Consejo de Seguridad Nuclear, Madrid (Spain))

    1991-01-01

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

  11. CONTROL OF POPULATION DENSITIES SURROUNDING NUCLEAR POWER PLANTS. VOLUME 5 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, jA.V.

    2010-01-01

    report on HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL,5 of HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, ANDHealth and Safety Impacts of Nuclear, Geo- thermal, and

  12. Risk Assessment in Support of DOE Nuclear Safety, Risk Information Notice, June 2010

    Broader source: Energy.gov [DOE]

    On August 12, 2009, the Defense Nuclear Facilities Safety Board(DNFSB) issued Recommendation 2009?1, Risk Assessment Methodologies at Defense Nuclear Facilities. Thisrecommendation focused on the...

  13. Facilities & Capabilities | Nuclear Science | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    R&D in radioisotopes and stable isotopes; radioisotope production; radiochemistry; global nuclear security technologies; fusion energy science; nuclear criticality safety; and all...

  14. Test Automation of SafetyCritical Reactive Systems J. Peleska \\Lambda M. Siegel y

    E-Print Network [OSTI]

    Peleska, Jan - Fachbereich 3

    of tests is only possible on the basis of formal specifications with well­defined semantics. JustTest Automation of Safety­Critical Reactive Systems J. Peleska \\Lambda M. Siegel y \\Lambda: mis@informatik.uni­kiel.de Abstract This article focuses on test automation for safety

  15. Pipeline Safety Our goal is to provide standard test methods and critical data to

    E-Print Network [OSTI]

    Pipeline Safety METALS Our goal is to provide standard test methods and critical data to the pipeline industry to improve safety and reliability. Of particular interest is the testing of high strength pipeline steels, which could enable higher volume gas transport and reduce energy costs. However

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

    SciTech Connect (OSTI)

    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.

  17. Nuclear Criticality as a Contributor to Gamma Ray Burst Events

    E-Print Network [OSTI]

    Robert Bruce Hayes

    2013-01-15

    Most gamma ray bursts are able to be explained using supernovae related phenomenon. Some measured results still lack compelling explanations and a contributory cause from nuclear criticality is proposed. This is shown to have general properties consistent with various known gamma ray burst properties. The galactic origin of fast rise exponential decay gamma ray bursts is considered a strong candidate for these types of events.

  18. Validation of MCNP6.1 for Criticality Safety of Pu-Metal, -Solution, and -Oxide Systems

    SciTech Connect (OSTI)

    Kiedrowski, Brian C.; Conlin, Jeremy Lloyd; Favorite, Jeffrey A.; Kahler, III, Albert C.; Kersting, Alyssa R.; Parsons, Donald K.; Walker, Jessie L.

    2014-05-13

    Guidance is offered to the Los Alamos National Laboratory Nuclear Criticality Safety division towards developing an Upper Subcritical Limit (USL) for MCNP6.1 calculations with ENDF/B-VII.1 nuclear data for three classes of problems: Pu-metal, -solution, and -oxide systems. A benchmark suite containing 1,086 benchmarks is prepared, and a sensitivity/uncertainty (S/U) method with a generalized linear least squares (GLLS) data adjustment is used to reject outliers, bringing the total to 959 usable benchmarks. For each class of problem, S/U methods are used to select relevant experimental benchmarks, and the calculational margin is computed using extreme value theory. A portion of the margin of sub criticality is defined considering both a detection limit for errors in codes and data and uncertainty/variability in the nuclear data library. The latter employs S/U methods with a GLLS data adjustment to find representative nuclear data covariances constrained by integral experiments, which are then used to compute uncertainties in keff from nuclear data. The USLs for the classes of problems are as follows: Pu metal, 0.980; Pu solutions, 0.973; dry Pu oxides, 0.978; dilute Pu oxide-water mixes, 0.970; and intermediate-spectrum Pu oxide-water mixes, 0.953.

  19. CTMCONTROL: Addressing the MC/DC Objective for Safety-Critical Automotive Software

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    , testing, MC/DC, ISO 26262, AUTOSAR. 1 INTRODUCTION ISO 26262 [1] is one of the main points of reference-critical software. AUTOSAR, since release 4.0 in 2011 [2], refers to the demands of ISO 26262 for safety A (most critical) software for airborne systems before they can receive an official approval/certification

  20. Nuclear Safety Information | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann JacksonDepartment ofOffice|in the subsurface isProject |NewsATVA Watts Bar Nuclear

  1. Nuclear safety | Princeton Plasma Physics Lab

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformationJesseworkSURVEY UNIVERSE The 2014 surveyNuclear and Particlesafety Subscribe to

  2. ORAU: Radiation and Nuclear Safety Services

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformationJesseworkSURVEY UNIVERSE The 2014 surveyNuclear andTwo-Phase75

  3. Uncertainties in the effects of burnup and their impact on criticality safety licensing criteria

    SciTech Connect (OSTI)

    Carlson, R.W.; Fisher, L.E.

    1990-07-13

    Current criteria for criticality safety for spent fuel shipping and storage casks are conservative because no credit is permitted for the effects of burnup of the fuel inside the cask. Cask designs that will transport and store large numbers of fuel assemblies (20 or more) must devote a substantial part of their payload to criticality control measures if they are to meet this criteria. The Department of Energy is developing the data necessary to support safety analyses that incorporate the effects of burnup for the next generation of spent fuel shipping casks. The efforts described here are devoted to the development of acceptance criteria that will be the basis for accepting safety analyses. Preliminary estimates of the uncertainties of the effects of burnup have been developed to provide a basis for the consideration of critically safety criteria. The criticality safety margins in a spent fuel shipping or storage cask are dominated by the portions of a fuel assembly that are in low power regions of a reactor core, and the reactor operating conditions are very different from spent fuel storage or transport cask conditions. Consequently, the experience that has been gathered during years of reactor operation does not apply directly to the prediction of criticality safety margins for spent fuel shipping or storage casks. The preliminary estimates of the uncertainties presented in this paper must be refined by both analytical and empirical studies that address both the magnitude of the uncertainties and their interdependence. 9 refs., 5 figs.

  4. METHODOLOGIES FOR REVIEW OF THE HEALTH AND SAFETY ASPECTS OF PROPOSED NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL SITES AND FACILITIES. VOLUME 9 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    considering the Nuclear Regulatory Commission safety reviewof the Nuclear Regulatory Commission safety review has onlyc..1 HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND

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

    SciTech Connect (OSTI)

    Selin, I.

    1994-12-31

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

  6. Double-clad nuclear-fuel safety rod

    DOE Patents [OSTI]

    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.

  7. CRAD, NNSA - Criticality Safety (CS) | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i n c i p a l DeInsulation at04-86)ContractorsCNGFact SLCriticality Safety (CS).

  8. Locations of criticality alarms and nuclear accident dosimeters at Hanford

    SciTech Connect (OSTI)

    Not Available

    1992-08-01

    Hanford facilities that contain fissionable materials capable of achieving critical mass are monitored with nuclear accident dosimeters (NADS) in compliance with the requirements of DOE Order 5480.11, Chapter XI, Section 4.c. (DOE 1988). The US Department of Energy (DOE) Richland Field Office (RL) has assigned the responsibility for maintaining and evaluating the Hanford NAD system to the Instrumentation and External Dosimetry (I ED) Section of Pacific Northwest Laboratory's (PNL's) Health Physics Department. This manual provides a description of the Hanford NAD, criteria and instructions for proper NAD placement, and the locations of these dosimeters onsite.

  9. Some aspects of nuclear power plant safety under war conditions

    SciTech Connect (OSTI)

    Stritar, A.; Mavko, B.; Susnik, J.; Sarler, B. (Jozef Stefan Inst., Ljubljana (Slovenia))

    1993-02-01

    In the summer of 1991, the Krsko nuclear power plant in Slovenia found itself in an area of military operations. This was probably the first commercial nuclear power plant to have been threatened by an attack by fighter jets. A number of never-before-asked questions had to be answered by the operating staff and supporting organizations. Some aspects of nuclear power plant safety under war conditions are described, such as the selection of the best plant operating state before the attack and the determination of plant system vulnerability and dose releases from the potentially damaged spent fuel in the spent-fuel pit. The best operating mode to which the plant should be brought before the attack is cold shutdown, and radiological consequences to the environment after the spent fuel is damaged and the water in the pit is lost are not very high. The problem of nuclear power plant safety under war conditions should be addressed in more detail in the future.

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

    SciTech Connect (OSTI)

    Kerr, C E

    2009-10-07

    The DOE Office of Enforcement expects LLNL to 'implement comprehensive management and independent assessments that are effective in identifying deficiencies and broader problems in safety and security programs, as well as opportunities for continuous improvement within the organization' and to 'regularly perform assessments to evaluate implementation of the contractor's processes for screening and internal reporting.' LLNL has a self-assessment program, described in ES&H Manual Document 4.1, that includes line, management and independent assessments. LLNL also has in place a process to identify and report deficiencies of nuclear, worker safety and health and security requirements. In addition, the DOE Office of Enforcement expects LLNL to evaluate 'issues management databases to identify adverse trends, dominant problem areas, and potential repetitive events or conditions' (page 14, DOE Enforcement Process Overview, December 2007). LLNL requires that all worker safety and health and nuclear safety noncompliances be tracked as 'deficiencies' in the LLNL Issues Tracking System (ITS). Data from the ITS are analyzed for worker safety and health (WSH) and nuclear safety noncompliances that may meet the threshold for reporting to the DOE Noncompliance Tracking System (NTS). This report meets the expectations defined by the DOE Office of Enforcement to review the assessments conducted by LLNL, analyze the issues and noncompliances found in these assessments, and evaluate the data in the ITS database to identify adverse trends, dominant problem areas, and potential repetitive events or conditions. The report attempts to answer three questions: (1) Is LLNL evaluating its programs and state of compliance? (2) What is LLNL finding? (3) Is LLNL appropriately managing what it finds? The analysis in this report focuses on data from the first quarter of 2008 (January through March). This quarter is analyzed within the context of information identified in previous quarters to include April 2007 through March 2008. The results from analyzing the deficiencies are presented in accordance with the two primary NTS reporting thresholds for WSH and nuclear safety noncompliances: (1) those related to certain events or conditions and (2) those that are management issues. In addition, WSH noncompliances were also analyzed to determine if any fell under the 'other significant condition' threshold. This report also identifies noncompliance topical areas that may have issues that do not meet the NTS reporting threshold but should remain under observation. These are placed on the 'watch list' for continued analysis.

  11. Covariance matrices for use in criticality safety predictability studies

    SciTech Connect (OSTI)

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

    1997-09-01

    Criticality predictability applications require as input the best available information on fissile and other nuclides. In recent years important work has been performed in the analysis of neutron transmission and cross-section data for fissile nuclei in the resonance region by using the computer code SAMMY. The code uses Bayes method (a form of generalized least squares) for sequential analyses of several sets of experimental data. Values for Reich-Moore resonance parameters, their covariances, and the derivatives with respect to the adjusted parameters (data sensitivities) are obtained. In general, the parameter file contains several thousand values and the dimension of the covariance matrices is correspondingly large. These matrices are not reported in the current evaluated data files due to their large dimensions and to the inadequacy of the file formats. The present work has two goals: the first is to calculate the covariances of group-averaged cross sections from the covariance files generated by SAMMY, because these can be more readily utilized in criticality predictability calculations. The second goal is to propose a more practical interface between SAMMY and the evaluated files. Examples are given for {sup 235}U in the popular 199- and 238-group structures, using the latest ORNL evaluation of the {sup 235}U resonance parameters.

  12. DICE: Database for the International Criticality Safety Benchmark Evaluation Program Handbook

    SciTech Connect (OSTI)

    Nouri, Ali [Organization for Economic Co-operation and Development Nuclear Energy Agency (France); Nagel, Pierre [Organization for Economic Co-operation and Development Nuclear Energy Agency (France); Briggs, J. Blair [Idaho National Engineering and Environmental Laboratory (United States); Ivanova, Tatiana [Institute of Physics and Power Engineering (Russian Federation)

    2003-09-15

    The 2002 edition of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments' (ICSBEP Handbook) spans more than 26 000 pages and contains 330 evaluations with benchmark specifications for 2881 critical or near-critical configurations. With such a large content, it became evident that the users needed more than a broad and qualitative classification of experiments to make efficient use of the ICSBEP Handbook. This paper describes the features of Database for the International Handbook of Evaluated Criticality Safety Benchmark Experiments (DICE), which is a database for the ICSBEP Handbook. The DICE program contains a relational database loaded with selected information from each configuration and a users' interface that enables one to query the database and to extract specific parameters. Summary descriptions of each experimental configuration can also be obtained. In addition, plotting capabilities provide the means of comparing neutron spectra and sensitivity coefficients for a set of configurations.

  13. Criticality safety evaluation of disposing of K Basin sludge in double-shell tank AW-105

    SciTech Connect (OSTI)

    ROGERS, C.A.

    1999-06-04

    A criticality safety evaluation is made of the disposal of K Basin sludge in double-shell tank (DST) AW-105 located in the 200 east area of Hanford Site. The technical basis is provided for limits and controls to be used in the development of a criticality prevention specification (CPS). A model of K Basin sludge is developed to account for fuel burnup. The iron/uranium mass ration required to ensure an acceptable magrin of subcriticality is determined.

  14. Criticality safety concerns of uranium deposits in cascade equipment

    SciTech Connect (OSTI)

    Plaster, M.J. [Lockheed Martin Utility Services, Inc., Piketon, OH (United States)

    1996-12-31

    The Paducah and Portsmouth Gaseous Diffusion Plants enrich uranium in the {sup 235}U isotope by diffusing gaseous uranium hexafluoride (UF{sub 6}) through a porous barrier. The UF{sub 6} gaseous diffusion cascade utilized several thousand {open_quotes}stages{close_quotes} of barrier to produce highly enriched uranium (HEU). Historically, Portsmouth has enriched the Paducah Gaseous Diffusion Plant`s product (typically 1.8 wt% {sup 235}U) as well as natural enrichment feed stock up to 97 wt%. Due to the chemical reactivity of UF{sub 6}, particularly with water, the formation of solid uranium deposits occur at a gaseous diffusion plant. Much of the equipment operates below atmospheric pressure, and deposits are formed when atmospheric air enters the cascade. Deposits may also be formed from UF{sub 6} reactions with oil, UF{sub 6} reactions with the metallic surfaces of equipment, and desublimation of UF{sub 6}. The major deposits form as a result of moist air in leakage due to failure of compressor casing flanges, blow-off plates, seals, expansion joint convolutions, and instrument lines. This report describes criticality concerns and deposit disposition.

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

    SciTech Connect (OSTI)

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

    2010-02-01

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

  16. Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels

    E-Print Network [OSTI]

    Chen, Sheng

    Walking and Climbing Service Robots for Safety Inspection of Nuclear Reactor Pressure Vessels B of Electronics and Computer Science, University of Southampton, Southampton, UK Abstract: Nuclear reactor and the usefulness of these robots for improving safety inspection of nuclear reactors in general are discussed

  17. ROBERT J. BUDNITZ Occupation: Physicist in Energy/Environmental Research and Nuclear Safety

    E-Print Network [OSTI]

    Ajo-Franklin, Jonathan

    ROBERT J. BUDNITZ Occupation: Physicist in Energy/Environmental Research and Nuclear Safety Birth December 2004 to September 2007 (in Livermore): Leader, Nuclear & Risk Science Group, Energy & Environment Directorate Associate Program Leader for Nuclear Systems Safety and Security, E&E Directorate October 2002

  18. Microsoft PowerPoint - Fire Safety workshop NQA-1 CGD 4 29 15...

    Office of Environmental Management (EM)

    Programs Workshop Fire Safety Track May 5th, 2015 Overview NQA 1 Commercial Grade Dedication Critical Characteristics Department of Energy Nuclear and Facility Safety Programs...

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

    SciTech Connect (OSTI)

    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.

  20. Price-Anderson Nuclear Safety Enforcement Program. 1997 annual report

    SciTech Connect (OSTI)

    NONE

    1998-01-01

    This report summarizes activities in the Department of Energy's Price-Anderson Amendments Act (PAAA) Enforcement Program in calendar year 1997 and highlights improvements planned for 1998. The DOE Enforcement Program involves the Office of Enforcement and Investigation in the DOE Headquarters Office of Environment, Safety and Health, as well as numerous PAAA Coordinators and technical advisors in DOE Field and Program Offices. The DOE Enforcement Program issued 13 Notices of Violation (NOV`s) in 1997 for cases involving significant or potentially significant nuclear safety violations. Six of these included civil penalties totaling $440,000. Highlights of these actions include: (1) Brookhaven National Laboratory Radiological Control Violations / Associated Universities, Inc.; (2) Bioassay Program Violations at Mound / EG and G, Inc.; (3) Savannah River Crane Operator Uptake / Westinghouse Savannah River Company; (4) Waste Calciner Worker Uptake / Lockheed-Martin Idaho Technologies Company; and (5) Reactor Scram and Records Destruction at Sandia / Sandia Corporation (Lockheed-Martin).

  1. DEZENT: A Safety-Critical Real-Time Approach Decentralized Electric Power Management

    E-Print Network [OSTI]

    Wedde, Horst F.

    for establishing technologies based on solar or wind power, or on renewable energy sources, is an adequate- selves to the producer/ consumer problems within a balance responsibility entity, i.e. a 10 kV power areaDEZENT: A Safety-Critical Real-Time Approach for Decentralized Electric Power Management H. F

  2. Testing of Safety-Critical Software Embedded in an Artificial Heart

    E-Print Network [OSTI]

    Testing of Safety-Critical Software Embedded in an Artificial Heart Sungdeok Cha1 , Sehun Jeong1 frequently to control medical devices such as artificial heart or robotic surgery system. While much (KAOC). It is a state-of-the-art artificial heart which completed animal testing phase. We per- formed

  3. Safety culture in the nuclear power industry : attributes for regulatory assessment

    E-Print Network [OSTI]

    Alexander, Erin L

    2004-01-01

    Safety culture refers to the attitudes, behaviors, and conditions that affect safety performance and often arises in discussions following incidents at nuclear power plants. As it involves both operational and management ...

  4. The importance of data in safety-critical systems Neil Storey

    E-Print Network [OSTI]

    Storey, Neil

    , industry-specific standards such as those in the civil aircraft, military, nuclear and railway sectors [3-6] give very little guidance in this area. One of the few industrial sectors that do have standards standards and guidelines relating to safety-related systems. Generic standards, such as IEC 61508 [2

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

    SciTech Connect (OSTI)

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

    2012-07-01

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

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

    SciTech Connect (OSTI)

    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.

  7. Office Of Nuclear Energy

    Broader source: Energy.gov (indexed) [DOE]

    Attributes of Software-Based Safety Critical Instrumentation and Control Systems in Nuclear Power Plants) (Carol Smidts) (The Ohio State University) (NEET 2) October 28-29, 2015...

  8. Institute for Critical Technology and Applied Science Seminar Series Emerging Technologies in Nuclear

    E-Print Network [OSTI]

    Crawford, T. Daniel

    Institute for Critical Technology and Applied Science Seminar Series Emerging Technologies in Nuclear Science & Engineering ­ Development of novel techniques/tools using particle transport theory methodologies with Alireza Haghighat, Nuclear Engineering Program, Mechanical Engineering Department Virginia

  9. Nonextensive critical effects in relativistic nuclear mean field models

    E-Print Network [OSTI]

    J. Rozynek; G. Wilk

    2011-02-22

    We present a possible extension of the usual relativistic nuclear mean field models widely used to describe nuclear matter towards accounting for the influence of possible intrinsic fluctuations caused by the environment. Rather than individually identifying their particular causes we concentrate on the fact that such effects can be summarily incorporated in the changing of the statistical background used, from the usual (extensive) Boltzman-Gibbs one to the nonextensive taken in the form proposed by Tsallis with a dimensionless nonextensivity parameter $q$ responsible for the above mentioned effects (for $q \\rightarrow 1$ one recovers the usual BG case). We illustrate this proposition on the example of the QCD-based Nambu - Jona-Lasinio (NJL) model of a many-body field theory describing the behavior of strongly interacting matter presenting its nonextensive version. We check the sensitivity of the usual NJL model to a departure from the BG scenario expressed by the value of $| q - 1|$, in particular in the vicinity of critical points.

  10. AN INTEGRATED ENVIRONMENT OF S/W SPECIFICATION AND V&V FOR SAFETY-CRITICAL SYSTEMS

    E-Print Network [OSTI]

    AN INTEGRATED ENVIRONMENT OF S/W SPECIFICATION AND V&V FOR SAFETY-CRITICAL SYSTEMS Abstract, an integrated environment of S/W specification and V&V is proposed for safety-critical systems. Integrated environment consists of SIS-RT for concept phase, NuSRS for requirement phase, NuSDS for design phase, and Nu

  11. Nuclear Safety R&D in the Waste Processing Technology Development...

    Office of Environmental Management (EM)

    Baione Office of Waste Processing DOE-EM Office of Engineering & Technology 2 Outline Nuclear Safety Research & Development Overview Summary of EM- NSR&D Presentations from...

  12. Nuclear Safety Risk Management in Refueling Outage of Qinshan Nuclear Power Plant

    SciTech Connect (OSTI)

    Meijing Wu; Guozhang Shen [Qinshan Nuclear power company (China)

    2006-07-01

    The NPP is used to planning maintenance, in-service inspection, surveillance test, fuel handling and design modification in the refueling outage; the operator response capability will be reduced plus some of the plant systems out of service or loss of power at this time. Based on 8 times refueling outage experiences of the Qinshan NPP, this article provide some good practice and lesson learned for the nuclear safety risk management focus at four safety function areas of Residual Heat Removal Capability, Inventory Control, Power availability and Reactivity control. (authors)

  13. INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

    SciTech Connect (OSTI)

    J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

    2008-09-01

    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. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

  14. NEW POSITIONS AT THE NEA DRAFT DESCRIPTIONS 1-Nuclear Safety Specialist Grade A3

    E-Print Network [OSTI]

    Psarrakos, Panayiotis

    , technological and legal bases required for a safe, environmentally friendly and economical use of nuclear energy in the OECD Nuclear Energy Agency (NEA). The Nuclear Energy Agency (NEA) was established to assist its member1 NEW POSITIONS AT THE NEA ­ DRAFT DESCRIPTIONS 1-Nuclear Safety Specialist ­ Grade A3 The NEA

  15. Nuclear Data Capabilities Supported by the DOE NCSP

    E-Print Network [OSTI]

    Danon, Yaron

    Nuclear Data Capabilities Supported by the DOE NCSP Symposium on Nuclear Data for Criticality responsible for developing, implementing, and maintaining nuclear criticality safety. 3 #12;NCSP Technical the Production Codes and Methods for Criticality Safety Engineers (e.g. MCNP, SCALE, & COG) · Nuclear Data

  16. Sandia Energy - Transportation Safety

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Transportation Safety Home Stationary Power Nuclear Fuel Cycle Nuclear Energy Safety Technologies Risk and Safety Assessment Transportation Safety Transportation SafetyTara...

  17. Critical technologies for reactors used in nuclear electric propulsion

    SciTech Connect (OSTI)

    Bhattacharyya, S.K. (Argonne National Lab., IL (United States))

    1993-01-01

    Nuclear electric Propulsion (NEP) systems are expected to play a significant role in the exploration and exploitation of space. Unlike nuclear thermal propulsion (NTP) systems in which the hot reactor coolant is directly discharged from nozzles to provide the required thrust, NEP systems include electric power generation and conditioning units that in turn are used to drive electric thrusters. These thrusters accelerate sub atomic particles to produce thrust. The major advantage of NEP systems is the ability to provide very high specific impulses ([approximately]5000 s) that minimize the requirement for propellants. In addition, the power systems used in NEP could pro vide the dual purpose of also providing power for the missions at the destination. This synergism can be exploited in shared development costs. The NEP systems produce significantly lower thrust that NTP systems and are generally more massive. Both systems have their appropriate roles in a balanced space program. The technology development needs of NEP systems differ in many important ways from the development needs for NTP systems because of the significant differences in the operating conditions of the systems. The NEP systems require long-life reactor power systems operating at power levels that are considerably lower than those for NTP systems. In contrast, the operational lifetime of an NEP system (years) is orders of magnitude longer than the operational lifetime of NTP systems (thousands of second). Thus, the critical issue of NEP is survivability and reliable operability for very long times at temperatures that are considerably more modest than the temperatures required for effective NTP operations but generally much higher than those experienced in terrestrial reactors.

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

    SciTech Connect (OSTI)

    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. A REVIEW OF AIR QUALITY MODELING TECHNIQUES. VOLUME 8 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Rosen, L.C.

    2010-01-01

    report on HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL,8 of HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, ANDHealth and Safety Impacts of Nuclear, Geo- thermal, and

  20. Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository - BMT1 of the DECOVALEX III project. Part 3: Effects of THM coupling in sparsely fractured rocks

    E-Print Network [OSTI]

    2004-01-01

    Paris, France e. Canadian Nuclear Safety Commission (CNSC),Technology (KTH), Canadian Nuclear Safety Commission (CNSC),order): the Canadian Nuclear Safety Commission; the

  1. Importance of Nuclear DataImportance of Nuclear Data to the Naval Nuclear Propulsion Program

    E-Print Network [OSTI]

    Danon, Yaron

    Importance of Nuclear DataImportance of Nuclear Data to the Naval Nuclear Propulsion Program Don of Evaluated Criticality Safety Benchmark Experiments," NEA/NSC/DOC(95)03, 2009. #12;MC21 Analysis of LEU COMP

  2. Nuclear electric propulsion operational reliability and crew safety study

    SciTech Connect (OSTI)

    Karns, J.J.; Fragola, J.R.; Kahan, L.; Pelaccio, D. (Science Applications International Corporation, 8 W 40th St., 14th Floor, New York, New York 10018 (United States))

    1993-01-20

    The central purpose of this analysis is to assess the achievability'' of a nuclear electric propulsion (NEP) system in a given mission. Achievability'' is a concept introduced to indicate the extent to which a system that meets or achieves its design goals might be implemented using the existing technology base. In the context of this analysis, the objective is to assess the achievability of an NEP system for a manned Mars mission as it pertains to operational reliability and crew safety goals. By varying design parameters, then examining the resulting system achievability, the design and mission risk drivers can be identified. Additionally, conceptual changes in design approach or mission strategy which are likely to improve overall achievability of the NEP system can be examined.

  3. Criticality safety strategy for the Fuel Cycle Facility electrorefiner at Argonne National Laboratory, West

    SciTech Connect (OSTI)

    Mariani, R.D.; Benedict, R.W. [Argonne National Lab., Idaho Falls, ID (United States); Lell, R.M.; Turski, R.B.; Fujita, E.K. [Argonne National Lab., IL (United States)

    1993-09-01

    The Integral Fast Reactor being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal-cooled reactors and a closed fuel cycle. Presently, the Fuel Cycle Facility (FCF) at ANL-West in Idaho Falls, Idaho is being modified to recycle spent metallic fuel from Experimental Breeder Reactor II as part of a demonstration project sponsored by the Department of Energy. A key component of the FCF is the electrorefiner (ER) in which the actinides are separated from the fission products. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt and refined uranium or uranium/plutonium products are deposited at cathodes. In this report, the criticality safety strategy for the FCF ER is summarized. FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. In order to show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOES) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOES, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that wig verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality.

  4. Enforcement handbook: Enforcement of DOE nuclear safety requirements

    SciTech Connect (OSTI)

    NONE

    1995-06-01

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

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

    SciTech Connect (OSTI)

    1983-02-01

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

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

    SciTech Connect (OSTI)

    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.

  7. RADIOLOGICAL EMERGENCY RESPONSE PLANNING FOR NUCLEAR POWER PLANTS IN CALIFORNIA. VOLUME 4 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Yen, W.W.S.

    2010-01-01

    Effect of Engineered Nuclear Safety, Volume 7, Number d I U.on Reactor Siting," Nuclear Safety, Vol. 7,No, 3, Springto Population," Nuclear Safety, Vol. 14, No.6, November-

  8. Forecast of criticality experiments and experimental programs needed to support nuclear operations in the United States of America: 1994--1999

    SciTech Connect (OSTI)

    Rutherford, D.

    1994-03-01

    This Forecast is generated by the Chair of the Experiment Needs Identification Workgroup (ENIWG), with input from Department of Energy and the nuclear community. One of the current concerns addressed by ENIWG was the Defense Nuclear Facilities Safety Board`s Recommendation 93-2. This Recommendation delineated the need for a critical experimental capability, which includes (1) a program of general-purpose experiments, (2) improving the information base, and (3) ongoing departmental programs. The nuclear community also recognizes the importance of criticality theory, which, as a stepping stone to computational analysis and safety code development, needs to be benchmarked against well-characterized critical experiments. A summary project of the Department`s needs with respect to criticality information includes (1) hands-on training, (2) criticality and nuclear data, (3) detector systems, (4) uranium- and plutonium-based reactors, and (5) accident analysis. The Workgroup has evaluated, prioritized, and categorized each proposed experiment and program. Transportation/Applications is a new category intended to cover the areas of storage, training, emergency response, and standards. This category has the highest number of priority-1 experiments (nine). Facilities capable of performing experiments include the Los Alamos Critical Experiment Facility (LACEF) along with Area V at Sandia National Laboratory. The LACEF continues to house the most significant collection of critical assemblies in the Western Hemisphere. The staff of this facility and Area V are trained and certified, and documentation is current. ENIWG will continue to work with the nuclear community to identify and prioritize experiments because there is an overwhelming need for critical experiments to be performed for basic research and code validation.

  9. Reliability Engineering and System Safety 92 (2007) 609618 The nuclear industry's transition to risk-informed regulation and

    E-Print Network [OSTI]

    2007-01-01

    , USA b Nuclear Power Engineering, Quality and Safety Management Department, Tokyo Electric Power improvement in safety based on Institute of Nuclear Power Operations (INPO) performance indicatorsReliability Engineering and System Safety 92 (2007) 609­618 The nuclear industry's transition

  10. Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications

    E-Print Network [OSTI]

    Heljanko, Keijo

    and control (I&C) systems play a crucial role in the operation of nuclear power plants (NPP) and other safety of the environment is covered. The reactor emergency cooling system is in use in an operating nuclear power plant is a reactor emergency cooling system in an operating nuclear power plant. 2. MODEL CHECKING METHODOLOGY

  11. Assessment of standard point-wise neutron data libraries for criticality safety analysis with a Monte Carlo code

    SciTech Connect (OSTI)

    Kolbe, E.; Vasiliev, A.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, CH 5232 Villigen PSI (Switzerland)

    2006-07-01

    This study addresses the assessment of standard continuous-energy neutron data libraries using the Monte Carlo radiation transport code MCNPX for light water reactor criticality safety applications based on a suite of low-enriched, thermal, compound uranium benchmarks and represents a continuation of previously performed analysis using the JEF-2.2 and JENDL-3.3 nuclear data libraries. The new work enhancing the previous study includes the application of the ENDF/B-6.8 neutron data library and employs the most recent official release of the code (MCNPX-2.5.0) with an improved S({alpha}, {beta}) thermal neutron scattering treatment. Particular attention is paid to the analysis of the spectrum-related characteristics of the modeled critical experimental configurations to define the range of applicability of the reported estimates of lower tolerance bounds for k{sub eff}. Inspection of trends in k{sub eff} versus the spectrum-related characteristics or design parameters has also been performed. (authors)

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

    SciTech Connect (OSTI)

    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 input: a serial interface with field equipment and a serial input from the FPD touch screen. The mechanism for data collection from the field equipment consists of the regular exchange of the data update request messages and target commands sent to the equipment and the update messages returned to the FPC. The data updates from field equipment control displays presented on the graphic pages. Touch screen contacts are decoded to identify physical position that was contacted. If that position corresponds with one of the buttons on the graphic page, the software uses that input to initiate the function defined for the particular button contacted. In this paper, the FPC will be illustrated as a standalone system as well as a module in a dedicated control system. The GDU design concepts and its design flow will be demonstrated. The dedication process of the QNX RTOS needed for the GDU will be highlighted. Finally, the GDU with a specific application example used in one of the nuclear power plants will be presented. (authors)

  13. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Testimony from the Nuclear Regulatory Commission indivi­ ofPlants", U.S. Nuclear Regulatory Commission Report WASH-Yellin, "The Nuclear Regulatory Commission's Reactor Safety

  14. Tornado vs. Hurricane Which is More Critical for Design of U.S. Nuclear Power Plants?

    Broader source: Energy.gov [DOE]

    Tornado vs. Hurricane Which is More Critical for Design of U.S. Nuclear Power Plants? Javad Moslemian Sargent & Lundy, LLC U. S. Department of Energy Natural Phenomena Hazards Meeting October 21-22, 2014

  15. 259NUCLEAR ENGINEERING AND TECHNOLOGY, VOL.38 NO.3 APRIL 2006 NuSEE: AN INTEGRATED ENVIRONMENT OF SOFTWARE

    E-Print Network [OSTI]

    . In safety-critical systems such as a Nuclear Power Plant (NPP), extremely high- confidence for software and analysis have become increasingly important in nuclear power plant (NPP) safety-critical systems. KEYWORDS : V&V, Software Specification, Safety-Critical System, Tool #12;260 NUCLEAR ENGINEERING

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  18. Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel

    SciTech Connect (OSTI)

    Rearden, B.T. [Oak Ridge National Laboratory (United States); Anderson, W.J. [Framatome ANP, Inc. (France); Harms, G.A. [Sandia National Laboratories (United States)

    2005-08-15

    Framatome ANP, Sandia National Laboratories (SNL), Oak Ridge National Laboratory (ORNL), and the University of Florida are cooperating on the U.S. Department of Energy Nuclear Energy Research Initiative (NERI) project 2001-0124 to design, assemble, execute, analyze, and document a series of critical experiments to validate reactor physics and criticality safety codes for the analysis of commercial power reactor fuels consisting of UO{sub 2} with {sup 235}U enrichments {>=}5 wt%. The experiments will be conducted at the SNL Pulsed Reactor Facility.Framatome ANP and SNL produced two series of conceptual experiment designs based on typical parameters, such as fuel-to-moderator ratios, that meet the programmatic requirements of this project within the given restraints on available materials and facilities. ORNL used the Tools for Sensitivity and Uncertainty Analysis Methodology Implementation (TSUNAMI) to assess, from a detailed physics-based perspective, the similarity of the experiment designs to the commercial systems they are intended to validate. Based on the results of the TSUNAMI analysis, one series of experiments was found to be preferable to the other and will provide significant new data for the validation of reactor physics and criticality safety codes.

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

    SciTech Connect (OSTI)

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

    1992-12-01

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

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

    SciTech Connect (OSTI)

    Forsberg, C.W.

    1990-07-01

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

  1. Fractality in momentum space: a signal of criticality in nuclear collisions

    E-Print Network [OSTI]

    Antoniou, Nikolaos G; Diakonos, Fotios K

    2015-01-01

    We show that critical systems of finite size develop a fractal structure in momentum space with anomalous dimension given in terms of the isotherm critical exponent delta of the corresponding infinite system. The associated power laws of transverse momentum correlations, in high-energy nuclear collisions, provide us with a signature of a critical point in strongly interacting matter according to the laws of QCD.

  2. Fractality in momentum space: a signal of criticality in nuclear collisions

    E-Print Network [OSTI]

    Nikolaos G. Antoniou; Nikolaos Davis; Fotios K. Diakonos

    2015-10-12

    We show that critical systems of finite size develop a fractal structure in momentum space with anomalous dimension given in terms of the isotherm critical exponent delta of the corresponding infinite system. The associated power laws of transverse momentum correlations, in high-energy nuclear collisions, provide us with a signature of a critical point in strongly interacting matter according to the laws of QCD.

  3. Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

    SciTech Connect (OSTI)

    J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

    2007-05-01

    Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June of 2003. The IRPhEP is patterned after its predecessor, the ICSBEP, but focuses on other integral measurements such as buckling, spectral characteristics, reactivity effects, reactivity coefficients, kinetics measurements, reaction-rate and power distributions, nuclide compositions and other miscellaneous types of measurements in addition to the critical configuration. The two projects are closely coordinated to avoid duplication of effort and to leverage limited resources to achieve a common goal. The purpose of the IRPhEP is to provide an extensively peer reviewed set of reactor physics related integral benchmark data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next generation reactors and establish the safety basis for operation of these reactors. While coordination and administration of the IRPhEP takes place at an international level, each participating country is responsible for the administration, technical direction, and priorities of the project within their respective countries. The work of the IRPhEP is documented in an OECD NEA Handbook entitled, “International Handbook of Evaluated Reactor Physics Benchmark Experiments.” The first edition of this Handbook, the 2006 Edition spans over 2000 pages and contains data from 16 different experimental series that were

  4. Integrated approach to nuclear materials safety management in the U.S. and Russia

    SciTech Connect (OSTI)

    Jardine, L.J.

    1997-06-01

    The United States and Russia are dismantling nuclear weapons and generating hundreds of tons of excess plutonium and high enriched uranium fissile nuclear materials that require disposition. The U.S. Department of Energy and the Ministry of the Russian Federation for Atomic Energy (Minatom) organizations are planning and implementing safe, secure storage and disposition operations for these materials in numerous facilities. This provides a new opportunity for technical exchanges between Russian and Western scientists that can establish an integrated and improved common safety culture for handling these materials. The development and use of personal relationships and joint projects among Russian and Western participants involved in fissile nuclear materials safety management contributes to improving nuclear materials nonproliferation and to making a safer world. Technical exchanges and workshops are being used to systematically identify opportunities in the nuclear fissile materials facilities to improve and ensure the safety of workers, the public, and the environment.

  5. Criticality safety validation: Simple geometry, single unit {sup 233}U systems

    SciTech Connect (OSTI)

    Putman, V.L.

    1997-06-01

    Typically used LMITCO criticality safety computational methods are evaluated for suitability when applied to INEEL {sup 233}U systems which reasonably can be modeled as simple-geometry, single-unit systems. Sixty-seven critical experiments of uranium highly enriched in {sup 233}U, including 57 aqueous solution, thermal-energy systems and 10 metal, fast-energy systems, were modeled. These experiments include 41 cylindrical and 26 spherical cores, and 41 reflected and 26 unreflected systems. No experiments were found for intermediate-neutron-energy ranges, or with interstitial non-hydrogenous materials typical of waste systems, mixed {sup 233}U and plutonium, or reflectors such as steel, lead, or concrete. No simple geometry experiments were found with cubic or annular cores, or approximating infinite sea systems. Calculations were performed with various tools and methodologies. Nine cross-section libraries, based on ENDF/B-IV, -V, or -VI.2, or on Hansen-Roach source data, were used with cross-section processing methods of MCNP or SCALE. The k{sub eff} calculations were performed with neutral-particle transport and Monte Carlo methods of criticality codes DANT, MCNP 4A, and KENO Va.

  6. DOE Cites Fluor Fernald Inc. for Nuclear Safety Violations |...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Secretary for Environment, Safety and Health The Price-Anderson Amendments Act of 1988 authorizes the Energy Department to undertake regulatory actions against contractors...

  7. Independent Oversight Assessment of the Nuclear Safety Culture...

    Office of Environmental Management (EM)

    the safety basis documents containing no mitigated accident scenarios. There are no criteriarequirements for addressing beyond design basis events. Accident binning is...

  8. CRAD, Nuclear Safety- Oak Ridge National Laboratory High Flux Isotope Reactor

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) used for a February 2007 assessment of the Nuclear Safety Program in preparation for restart of the Oak Ridge National Laboratory High Flux Isotope Reactor.

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

    SciTech Connect (OSTI)

    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.

  10. ACCELERATED TESTING OF NEUTRON-ABSORBING ALLOYS FOR NUCLEAR CRITICALITY CONTROL

    SciTech Connect (OSTI)

    Ronald E. Mizia

    2011-10-01

    The US Department of Energy requires nuclear criticality control materials be used for storage of highly enriched spent nuclear fuel used in government programs and the storage of commercial spent nuclear fuel at the proposed High-Level Nuclear Waste Geological Repository located at Yucca Mountain, Nevada. Two different metallic alloys (Ni-Cr-Mo-Gd and borated stainless steel) have been chosen for this service. An accelerated corrosion test program to validate these materials for this application is described and a performance comparison is made.

  11. Critical Behavior in Light Nuclear Systems: Experimental Aspects

    E-Print Network [OSTI]

    Y. G. Ma; J. B. Natowitz; R. Wada; K. Hagel; J. Wang; T. Keutgen; Z. Majka; M. Murray; L. Qin; P. Smith; R. Alfaro; J. Cibor; M. Cinausero; Y. El Masri; D. Fabris; E. Fioretto; A. Keksis; M. Lunardon; A. Makeev; N. Marie; E. Martin; A. Martinez-Davalos; A. Menchaca-Rocha; G. Nebbia; G. Prete; V. Rizzi; A. Ruangma; D. V. Shetty; G. Souliotis; P. Staszel; M. Veselsky; G. Viesti; E. M. Winchester; S. J. Yennello

    2005-04-14

    An extensive experimental survey of the features of the disassembly of a small quasi-projectile system with $A \\sim$ 36, produced in the reactions of 47 MeV/nucleon $^{40}$Ar + $^{27}$Al, $^{48}$Ti and $^{58}$Ni, has been carried out. Nuclei in the excitation energy range of 1-9 MeV/u have been investigated employing a new method to reconstruct the quasi-projectile source. At an excitation energy $\\sim$ 5.6 MeV/nucleon many observables indicate the presence of maximal fluctuations in the de-excitation processes. The fragment topological structure shows that the rank sorted fragments obey Zipf's law at the point of largest fluctuations providing another indication of a liquid gas phase transition. The caloric curve for this system shows a monotonic increase of temperature with excitation energy and no apparent plateau. The temperature at the point of maximal fluctuations is $8.3 \\pm 0.5$ MeV. Taking this temperature as the critical temperature and employing the caloric curve information we have extracted the critical exponents $\\beta$, $\\gamma$ and $\\sigma$ from the data. Their values are also consistent with the values of the universality class of the liquid gas phase transition. Taken together, this body of evidence strongly suggests a phase change in an equilibrated mesoscopic system at, or extremely close to, the critical point.

  12. Safety of Department of Energy-Owned Nuclear Reactors

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1986-09-23

    To establish reactor safety program requirements assure that the safety of each Department of Energy-owned (DOE-owned) reactor is properly analyzed, evaluated, documented, and approved by DOE; and reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate protection for health and safety and will be in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. Cancels Chap. 6 of DOE O 5480.1A. Paragraphs 7b(3), 7e(3) & 8c canceled by DOE O 5480.23 & canceled by DOE N 251.4 of 9-29-95.

  13. Energy Department and Catholic University Improve Safety of Nuclear Waste

    Broader source: Energy.gov [DOE]

    A new waste processing plant in Washington will help to safely remove nuclear and chemical waste, thanks to research from Catholic University.

  14. Nuclear Safety Workshop Agenda - Post Fukushima Initiatives and...

    Broader source: Energy.gov (indexed) [DOE]

    Agenda Post Fukushima Initiatives and Results In response to the March 2011 accident at the Fukushima Daiichi nuclear power plant, Secretary Chu initiated a series of actions to...

  15. Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel

    SciTech Connect (OSTI)

    Radulescu, Georgeta; Mueller, Don; Goluoglu, Sedat; Hollenbach, Daniel F; Fox, Patricia B

    2007-10-01

    The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform postclosure criticality calculations. The validation process applies the criticality analysis methodology approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report. The application systems for this validation consist of waste packages containing transport, aging, and disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types, initial enrichments, and burnup values that are expected from the waste stream and of varying degree of internal component degradation that may occur over the 10,000-year regulatory time period. The criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP) transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit (LBTL) functions for CSNF waste packages are determined based on the guidance in ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1. The development of this report is consistent with Test Plan for: Range of Applicability and Bias Determination for Postclosure Criticality. This calculation report has been developed in support of licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual TAD canister.

  16. Composite neutron absorbing coatings for nuclear criticality control

    DOE Patents [OSTI]

    Wright, Richard N.; Swank, W. David; Mizia, Ronald E.

    2005-07-19

    Thermal neutron absorbing composite coating materials and methods of applying such coating materials to spent nuclear fuel storage systems are provided. A composite neutron absorbing coating applied to a substrate surface includes a neutron absorbing layer overlying at least a portion of the substrate surface, and a corrosion resistant top coat layer overlying at least a portion of the neutron absorbing layer. An optional bond coat layer can be formed on the substrate surface prior to forming the neutron absorbing layer. The neutron absorbing layer can include a neutron absorbing material, such as gadolinium oxide or gadolinium phosphate, dispersed in a metal alloy matrix. The coating layers may be formed by a plasma spray process or a high velocity oxygen fuel process.

  17. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Risks In U.S. Commercial Nuclear Power Plants", U.S. NuclearCommission, "The, Safety of Nuclear Power Reactors (Light-October 1, 1976. "Nuclear Power and the Environment," a

  18. The Nuclear Education and Staffing Challenge: Rebuilding Critical Skills in Nuclear Science and Technology

    SciTech Connect (OSTI)

    Wogman, Ned A.; Bond, Leonard J.; Waltar, Alan E.; Leber, R E.

    2005-01-01

    The United States, the Department of Energy (DOE) and its National Laboratories, including the Pacific Northwest National Laboratory (PNNL), are facing a serious attrition of nuclear scientists and engineers and their capabilities through the effects of aging staff. Within the DOE laboratories, 75% of nuclear personnel will be eligible to retire by 2010. It is expected that there will be a significant loss of senior nuclear science and technology staff at PNNL within five years. PNNL's nuclear legacy is firmly rooted in the DOE Hanford site, the World War II Manhattan Project, and subsequent programs. Historically, PNNL was a laboratory were 70% of its activities were nuclear/radiological, and now just under 50% of its current business science and technology are nuclear and radiologically oriented. Programs in the areas of Nuclear Legacies, Global Security, Nonproliferation, Homeland Security and National Defense, Radiobiology and Nuclear Energy still involve more than 1,000 of the 3,800 current laboratory staff, and these include more than 420 staff who are certified as nuclear/radiological scientists and engineers. This paper presents the current challenges faced by PNNL that require an emerging strategy to solve the nuclear staffing issues through the maintenance and replenishment of the human nuclear capital needed to support PNNL nuclear science and technology programs.

  19. The Nuclear Education and Staffing Challenge: Rebuilding Critical Skills in Nuclear Science and Technology.

    SciTech Connect (OSTI)

    Wogman, Ned A.; Bond, Leonard J.; Waltar, Alan E.; Leber, R. E.

    2005-01-01

    The United States, the Department of Energy (DOE) and its National Laboratories, including the Pacific Northwest National Laboratory (PNNL), are facing a serious attrition of nuclear scientists and engineers and their capabilities through the effects of aging staff. Within the DOE laboratories, 75% of nuclear personnel will be eligible to retire by 2010. It is expected that there will be a significant loss of senior nuclear science and technology staff at PNNL within five years. PNNL's nuclear legacy is firmly rooted in the DOE Hanford site, the World War II Manhattan Project, and subsequent programs. Historically, PNNL was a laboratory where 70% of its activities were nuclear/radiological, and now just under 50% of its current business science and technology are nuclear and radiologically oriented. Programs in the areas of Nuclear Legacies, Global Security, Nonproliferation, Homeland Security and National Defense, Radiobiology and Nuclear Energy still involve more than 1,000 of the 3,800 current laboratory staff, and these include more than 420 staff who are certified as nuclear/radiological scientists and engineers. This paper presents the current challenges faced by PNNL that require an emerging strategy to solve the nuclear staffing issues through the maintenance and replenishment of the human nuclear capital needed to support PNNL nuclear science and technology programs.

  20. Emergency preparedness source term development for the Office of Nuclear Material Safety and Safeguards-Licensed Facilities

    SciTech Connect (OSTI)

    Sutter, S.L.; Mishima, J.; Ballinger, M.Y.; Lindsey, C.G.

    1984-08-01

    In order to establish requirements for emergency preparedness plans at facilities licensed by the Office of Nuclear Materials Safety and Safeguards, the Nuclear Regulatory Commission (NRC) needs to develop source terms (the amount of material made airborne) in accidents. These source terms are used to estimate the potential public doses from the events, which, in turn, will be used to judge whether emergency preparedness plans are needed for a particular type of facility. Pacific Northwest Laboratory is providing the NRC with source terms by developing several accident scenarios for eleven types of fuel cycle and by-product operations. Several scenarios are developed for each operation, leading to the identification of the maximum release considered for emergency preparedness planning (MREPP) scenario. The MREPP scenarios postulated were of three types: fire, tornado, and criticality. Fire was significant at oxide fuel fabrication, UF/sub 6/ production, radiopharmaceutical manufacturing, radiopharmacy, sealed source manufacturing, waste warehousing, and university research and development facilities. Tornadoes were MREPP events for uranium mills and plutonium contaminated facilities, and criticalities were significant at nonoxide fuel fabrication and nuclear research and development facilities. Techniques for adjusting the MREPP release to different facilities are also described.

  1. Concentration of Actinides in Plant Mounds at Safety Test Nuclear Sites in Nevada

    SciTech Connect (OSTI)

    David S. Shafer; Jenna Gommes

    2008-09-15

    Plant mounds or blow-sand mounds are accumulations of soil particles and plant debris around large shrubs and are common features in deserts in the southwestern United States. Believed to be an important factor in their formation, the shrubs create surface roughness that causes wind-suspended particles to be deposited and resist further suspension. Shrub mounds occur in some plant communities on the Nevada Test Site, the Nevada Test and Training Range (NTTR), and Tonopah Test Range (TTR), including areas of surface soil contamination from past nuclear testing. In the 1970s as part of early studies to understand properties of actinides in the environment, the Nevada Applied Ecology Group (NAEG) examined the accumulation of isotopes of Pu, {sup 241}Am, and U in plant mounds at safety test sites. The NAEG studies found concentrations of these contaminants to be greater in shrub mounds than in the surrounding areas of desert pavement. For example, at Project 57 on the NTTR, it was estimated that 15 percent of the radionuclide inventory of the site was associated with shrub mounds, which accounted for 17 percent of the surface area of the site, a ratio of inventory to area of 0.85. At Clean Slate III at the TTR, 29 percent of the inventory was associated with approximately 32 percent of the site covered by shrub mounds, a ratio of 0.91. While the total inventory of radionuclides in intershrub areas was greater, the ratio of radionuclide inventory to area was 0.40 and 0.38, respectively, at the two sites. The comparison between the shrub mounds and adjacent desert pavement areas was made for only the top 5 cm since radionuclides at safety test sites are concentrated in the top 5 cm of intershrub areas. Not accounting for radionuclides associated with the shrub mounds would cause the inventory of contaminants and potential exposure to be underestimated. As part of its Environmental Restoration Soils Subproject, the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office has proposed that the majority of its contaminated soil 'Corrective Action Units', including the safety test sites, be closed by fencing and posting with administrative controls. The concentration of actinides in the shrub mounds has important implications for postclosure management of the safety test sites. Because resuspension factors at safety test sites can be three to four orders-of-magnitude higher than soil sites associated with atmospheric tests where criticality occurred, the shrub mounds are an important factor in stabilization of actinide contaminants. Loss of shrubs associated with mounds from fire or plant die-back from drought could cause radionuclides at these sites to become more prone to suspension and water erosion until the sites are stabilized. Alternatively, although shrub mounds are usually composed of predominantly fine sand size particles, smaller silt and clay size particles in them are often high in CaCO{sub 3} content. The CaCO{sub 3} may act as a cementing agent to limit erosion of the shrub mounds even if the vegetation cover is temporarily lost.

  2. Understanding IV&V in a Safety Critical and Complex Evolutionary Environment: The NASA Space Shuttle Program1

    E-Print Network [OSTI]

    Zelkowitz, Marvin V.

    Understanding IV&V in a Safety Critical and Complex Evolutionary Environment: The NASA Space, Maryland 20740, USA +1-301-403-8971 irus@fc-md.umd.edu 1 This work has been performed as NASA Subcontract No. 93-393B-FUSA from the NASA/IVV facility in Fairmont, WV to the Fraunhofer Center, Maryland

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

    SciTech Connect (OSTI)

    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.

  4. Safety-related issues associated with implementing wireless systems in nuclear facilities

    SciTech Connect (OSTI)

    Kaldenbach, B. J.; Ewing, P. D.; Moore, M. R.; Korsah, K. [Oak Ridge National Laboratory, MS-6006, P. O. Box 2008, Oak Ridge, TN 37831 (United States); Antonescu, C. E.; Govan, T. V. [U.S. Nuclear Regulatory Commission, MS T-10-D20, 11545 Rockville Pike, Rockville, MD 20852 (United States)

    2006-07-01

    Currently, wireless technology is not used as an integral element of safety-related systems in nuclear facilities. The most prevalent introductory uses of wireless technology are for in-facility communications among personnel and for supplemental information transmission. However, further system upgrades and implementations at new facilities might introduce wireless communications into safety-significant applications. This paper documents the deployment issues and implementation considerations that can contribute to the technical basis for guidance on wireless systems. (authors)

  5. Critical temperature for the nuclear liquid-gas phase transition (from multifragmentation and fission)

    E-Print Network [OSTI]

    V. A. Karnaukhov; H. Oeschler; A. Budzanowski; S. P. Avdeyev; A. S. Botvina; E. A. Cherepanov; W. Karcz; V. V. Kirakosyan; P. A. Rukoyatkin; I. Skwirczynska; E. Norbeck

    2008-01-29

    Critical temperature Tc for the nuclear liquid-gas phase transition is stimated both from the multifragmentation and fission data. In the first case,the critical temperature is obtained by analysis of the IMF yields in p(8.1 GeV)+Au collisions within the statistical model of multifragmentation (SMM). In the second case, the experimental fission probability for excited 188Os is compared with the calculated one with Tc as a free parameter. It is concluded for both cases that the critical temperature is higher than 16 MeV.

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

    SciTech Connect (OSTI)

    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.

  7. Development of Gd-Enriched Alloys for Spent Nuclear Fuel Applications--Part 1: Preliminary Characterization

    E-Print Network [OSTI]

    DuPont, John N.

    composition for any Gd level. Keywords gadolinium, neutron absorbing material, nuclear criticality safety support, (2) spent nuclear fuel geometry control, and (3) nuclear criticality safety. In additionDevelopment of Gd-Enriched Alloys for Spent Nuclear Fuel Applications--Part 1: Preliminary

  8. Technical Basis for U. S. Department of Energy Nuclear Safety Policy, DOE Policy 420.1

    Broader source: Energy.gov [DOE]

    This document provides the technical basis for the Department of Energy (DOE) Policy (P) 420.1, Nuclear Safety Policy, dated 2-8-2011. It includes an analysis of the revised Policy to determine whether it provides the necessary and sufficient high-level expectations that will lead DOE to establish and implement appropriate requirements to assure protection of the public, workers, and the environment from the hazards of DOE’s operation of nuclear facilities.

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

    SciTech Connect (OSTI)

    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.

  10. Space nuclear safety program: Progress report, April-June 1987

    SciTech Connect (OSTI)

    George, T.G. (comp.)

    1988-07-01

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

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

    SciTech Connect (OSTI)

    George, T.G. (comp.)

    1989-02-01

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

  12. Opportunities for improving regulations governing the seismic safety of large nuclear installations

    Office of Energy Efficiency and Renewable Energy (EERE)

    Opportunities for Improving Regulations Governing the Seismic Safety of Large Nuclear Installations Robert J. Budnitz, Ph.D. LBNL University of California, Berkeley, CA 94720 Andrew S. Whittaker, Ph.D., S.E. MCEER University at Buffalo, Buffalo, NY 14260

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

    SciTech Connect (OSTI)

    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.

  14. Superconducting Magnet Safety Nuclear Magnetic Resonance (NMR) facilities present unique hazards not found in most

    E-Print Network [OSTI]

    Maroncelli, Mark

    Superconducting Magnet Safety Nuclear Magnetic Resonance (NMR) facilities present unique hazards not found in most laboratory environments. The NMR facilities maintain superconducting magnets which have the units. Facility design and installation: Design and installation of a new NMR facility requires a number

  15. Numerical study of the thm effects on the near-field safety of a hypothetical nuclear waste repository - bmt1 of the decovalex iii project. part 1: conceptualization and characterization of the problems and summary of results

    E-Print Network [OSTI]

    2004-01-01

    mchiji@hazama.co.jp b. Canadian Nuclear Safety Commission (THM processes on the safety of nuclear waste repositories.coupling on the safety assessment of a nuclear fuel waste

  16. POWER PLANT RELIABILITY-AVAILABILITY AND STATE REGULATION. VOLUME 7 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    No. 75-50 (Ref. 1). Nuclear Safety, September 1975 to Augustreport on HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL,7 of HEAL TH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND

  17. AUDIT REPORT Follow-up on Nuclear Safety: Safety Basis and Quality Assurance at the Los Alamos National

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i n c i p a l De p u t y A s sconveyance of9, 2013ATVMEnergy'sNuclear Safety:

  18. Lessons learnt from ITER safety & licensing for DEMO and future nuclear fusion facilities

    E-Print Network [OSTI]

    Taylor, Neill

    2013-01-01

    One of the strong motivations for pursuing the development of fusion energy is its potentially low environmental impact and very good safety performance. But this safety and environmental potential can only be fully realized by careful design choices. For DEMO and other fusion facilities that will require nuclear licensing, S&E objectives and criteria should be set at an early stage and taken into account when choosing basic design options and throughout the design process. Studies in recent decades of the safety of fusion power plant concepts give a useful basis on which to build the S&E approach and to assess the impact of design choices. The experience of licensing ITER is of particular value, even though there are some important differences between ITER and DEMO. The ITER project has developed a safety case, produced a preliminary safety report and had it examined by the French nuclear safety authorities, leading to the licence to construct the facility. The key technical issues that arose during ...

  19. Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis

    SciTech Connect (OSTI)

    Enercon Services, Inc.

    2011-03-14

    Enercon Services, Inc. (ENERCON) was requested under Task Order No.2 to identify scientific and technical data needed to benchmark and justify Full Burnup Credit, which adds 16 fission products and 4 minor actinides1 to Actinide-Only burnup credit. The historical perspective for Full Burnup Credit is discussed, and interviews of organizations participating in burnup credit activities are summarized as a basis for identifying additional data needs and making recommendation. Input from burnup credit participants representing two segments of the commercial nuclear industry is provided. First, the Electric Power Research Institute (EPRI) has been very active in the development of Full Burnup Credit, representing the interests of nuclear utilities in achieving capacity gains for storage and transport casks. EPRI and its utility customers are interested in a swift resolution of the validation issues that are delaying the implementation of Full Burnup Credit [EPRI 2010b]. Second, used nuclear fuel storage and transportation Cask Vendors favor improving burnup credit beyond Actinide-Only burnup credit, although their discussion of specific burnup credit achievements and data needs was limited citing business sensitive and technical proprietary concerns. While Cask Vendor proprietary items are not specifically identified in this report, the needs of all nuclear industry participants are reflected in the conclusions and recommendations of this report. In addition, Oak Ridge National Laboratory (ORNL) and Sandia National Laboratory (SNL) were interviewed for their input into additional data needs to achieve Full Burnup Credit. ORNL was very open to discussions of Full Burnup Credit, with several telecoms and a visit by ENERCON to ORNL. For many years, ORNL has provided extensive support to the NRC regarding burnup credit in all of its forms. Discussions with ORNL focused on potential resolutions to the validation issues for the use of fission products. SNL was helpful in ENERCON's understanding of the difficult issues related to obtaining and analyzing additional cross section test data to support Full Burnup Credit. A PIRT (Phenomena Identification and Ranking Table) analysis was performed by ENERCON to evaluate the costs and benefits of acquiring different types of nuclear data in support of Full Burnup Credit. A PIRT exercise is a formal expert elicitation process with the final output being the ranking tables. The PIRT analysis (Table 7-4: Results of PIRT Evaluation) showed that the acquisition of additional Actinide-Only experimental data, although beneficial, was associated with high cost and is not necessarily needed. The conclusion was that the existing Radiochemical Assay (RCA) data plus the French Haut Taux de Combustion (HTC)2 and handbook Laboratory Critical Experiment (LCE) data provide adequate benchmark validation for Actinide-Only Burnup Credit. The PIRT analysis indicated that the costs and schedule to obtain sufficient additional experimental data to support the addition of 16 fission products to Actinide-Only Burnup Credit to produce Full Burnup Credit are quite substantial. ENERCON estimates the cost to be $50M to $100M with a schedule of five or more years. The PIRT analysis highlights another option for fission product burnup credit, which is the application of computer-based uncertainty analyses (S/U - Sensitivity/Uncertainty methodologies), confirmed by the limited experimental data that is already available. S/U analyses essentially transform cross section uncertainty information contained in the cross section libraries into a reactivity bias and uncertainty. Recent work by ORNL and EPRI has shown that a methodology to support Full Burnup Credit is possible using a combination of traditional RCA and LCE validation plus S/U validation for fission product isotopics and cross sections. Further, the most recent cross section data (ENDF/B-VII) can be incorporated into the burnup credit codes at a reasonable cost compared to the acquisition of equivalent experimental data. ENERCON concludes that even with the cos

  20. Enforcement Regulations and Directives - Nuclear Safety | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum12, 2015 Infographiclighbulbs - high-resolution JPG20,1LLCEnergy Nuclear

  1. FAQS Job Task Analyses - Nuclear Safety Specialist | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum12, 2015Executive Order 13514Construction Management FAQS JobRestorationofNuclear

  2. FTCP Site Specific Information - Chief of Nuclear Safety | Department of

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum12, 2015Executive Order14, 20111, 2015 FTCPDepartmentEnergy Chief of Nuclear

  3. Message from Chief of Nuclear Safety | Department of Energy

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties - WAPAEnergy May 28 WebinarProtectMessage from Chief of Nuclear

  4. 2012 Nuclear Safety Workshop Photos | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative FuelsofProgram:Y-12 Beta-3 Racetracks2 DOE Sustainability Awards2 National2012 Nuclear

  5. Nuclear Safety Policy, Guidance & Reports | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuelsof EnergyApril 2014Department ofWind CareerEnergy Nuclear

  6. OECD/NEA expert group on uncertainty analysis for criticality safety assessment: Results of benchmark on sensitivity calculation (phase III)

    SciTech Connect (OSTI)

    Ivanova, T.; Laville, C. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay aux Roses (France); Dyrda, J. [Atomic Weapons Establishment AWE, Aldermaston, Reading, RG7 4PR (United Kingdom); Mennerdahl, D. [E Mennerdahl Systems EMS, Starvaegen 12, 18357 Taeby (Sweden); Golovko, Y.; Raskach, K.; Tsiboulia, A. [Inst. for Physics and Power Engineering IPPE, 1, Bondarenko sq., 249033 Obninsk (Russian Federation); Lee, G. S.; Woo, S. W. [Korea Inst. of Nuclear Safety KINS, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Bidaud, A.; Sabouri, P. [Laboratoire de Physique Subatomique et de Cosmologie LPSC, CNRS-IN2P3/UJF/INPG, Grenoble (France); Patel, A. [U.S. Nuclear Regulatory Commission (NRC), Washington, DC 20555-0001 (United States); Bledsoe, K.; Rearden, B. [Oak Ridge National Laboratory ORNL, M.S. 6170, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Gulliford, J.; Michel-Sendis, F. [OECD/NEA, 12, Bd des Iles, 92130 Issy-les-Moulineaux (France)

    2012-07-01

    The sensitivities of the k{sub eff} eigenvalue to neutron cross sections have become commonly used in similarity studies and as part of the validation algorithm for criticality safety assessments. To test calculations of the sensitivity coefficients, a benchmark study (Phase III) has been established by the OECD-NEA/WPNCS/EG UACSA (Expert Group on Uncertainty Analysis for Criticality Safety Assessment). This paper presents some sensitivity results generated by the benchmark participants using various computational tools based upon different computational methods: SCALE/TSUNAMI-3D and -1D, MONK, APOLLO2-MORET 5, DRAGON-SUSD3D and MMKKENO. The study demonstrates the performance of the tools. It also illustrates how model simplifications impact the sensitivity results and demonstrates the importance of 'implicit' (self-shielding) sensitivities. This work has been a useful step towards verification of the existing and developed sensitivity analysis methods. (authors)

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

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA); Hui, Marvin M. (Sunnyvale, CA); Berglund, Robert C. (Saratoga, CA)

    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.

  8. 2012 Nuclear Safety Workshop Presentations | Department of Energy

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i n c i p a l De p u t y A s s i s tDistributedDISCLAIMER ThisNationalNuclear9 -

  9. NNSA Cites Los Alamos National Laboratory For Nuclear Safety Violations |

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformationJessework usesof Energy Moving Basic NERSCKeyNuclearBudget Presentations

  10. International Meeting on "Best Estimate" Methods in Nuclear Installation Safety Analysis (BE-2000) Washington, DC, November, 2000.

    E-Print Network [OSTI]

    Kunz, Robert Francis

    International Meeting on "Best Estimate" Methods in Nuclear Installation Safety Analysis (BE-2000 conditioning and analysis techniques has been undertaken which focuses on their relevance to nuclear reactor, to be used in United States Nuclear Regulatory Commission (USNRC) advanced T/H code consolidation efforts

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

    SciTech Connect (OSTI)

    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.

  12. Safety aspects of ground testing for large nuclear rockets

    SciTech Connect (OSTI)

    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.

  13. Criticality Safety Evaluation Report for the Cold Vacuum Drying (CVD) Facilities Process Water Handling System

    SciTech Connect (OSTI)

    KESSLER, S.F.

    2000-08-10

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified.

  14. Stakeholder Transportation Scorecard: Reviewing Nevada's Recommendations for Enhancing the Safety and Security of Nuclear Waste Shipments - 13518

    SciTech Connect (OSTI)

    Dilger, Fred C.; Ballard, James D.; Halstead, Robert J.

    2013-07-01

    As a primary stakeholder in the Yucca Mountain program, the state of Nevada has spent three decades examining and considering national policy regarding spent nuclear fuel and high-level radioactive waste transportation. During this time, Nevada has identified 10 issues it believes are critical to ensuring the safety and security of any spent nuclear fuel transportation program, and achieving public acceptance. These recommendations are: 1) Ship the oldest fuel first; 2) Ship mostly by rail; 3) Use dual-purpose (transportable storage) casks; 4) Use dedicated trains for rail shipments; 5) Implement a full-scale cask testing program; 6) Utilize a National Environmental Policy Act (NEPA) process for the selection of a new rail spur to the proposed repository site; 7) Implement the Western Interstate Energy Board (WIEB) 'straw man' process for route selection; 8) Implement Section 180C assistance to affected States, Tribes and localities through rulemaking; 9) Adopt safety and security regulatory enhancements proposed states; and 10) Address stakeholder concerns about terrorism and sabotage. This paper describes Nevada's proposals in detail and examines their current status. The paper describes the various forums and methods by which Nevada has presented its arguments and sought to influence national policy. As of 2012, most of Nevada's recommendations have been adopted in one form or another, although not yet implemented. If implemented in a future nuclear waste program, the State of Nevada believes these recommendations would form the basis for a successful national transportation plan for shipments to a geologic repository and/or centralized interim storage facility. (authors)

  15. Nuclear Resonance Fluorescence for Nuclear Materials Assay

    E-Print Network [OSTI]

    Quiter, Brian Joseph

    2010-01-01

    130] International Nuclear Safety Center, Available onlinefrom Inter- national Nuclear Safety Center (INSC) website(from International Nuclear Safety Center (INSC) website(

  16. Architectures for Cyber-Security Incident Reporting in Safety-Critical Systems Chris W. Johnson,

    E-Print Network [OSTI]

    Johnson, Chris

    of air navigation services and equipment, and maintenance organisations have the necessary procedures and policies for voluntary reporting of events that could affect aviation safety" (ICAO Resolution A32-15: ICAO Global Aviation Safety Plan) International support for voluntary incident reporting systems includes near

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  19. NEW - DOE O 420.1 Chg 1, Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. This Page Change is limited in scope to changes necessary to invoke DOE-STD-1104, Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Document, and revised DOE-STD-3009-2014, Preparation of Nonreactor Nuclear Facility Documented Safety Analysis as required methods. DOE O 420.1C Chg 1, dated 2-27-15, cancels DOE O 420.1C, dated 12-4-12.

  20. The Role of Independent Verification and Validation in Maintaining a Safety Critical Evolutionary Software in a

    E-Print Network [OSTI]

    Zelkowitz, Marvin V.

    Hartwick Road, Suite 500 College Park, Maryland 20740 irus@fc-md.umd.edu +1-301-403-8971 1 This work has with requirements for high reliability and mission safety taxes current development methods. In this paper we

  1. Questioning the Role of Requirements Engineering in the Causes of Safety-Critical Software Failures

    E-Print Network [OSTI]

    Johnson, C.W.

    Johnson,C.W. Holloway,C.M. T. Kelly (ed.), The First IET International Conference on System Safety, Institute of Engineering and Technology, Savoy Place, London, 6-8th June 2006 pp 352-360 IEE

  2. The potential role of new technology for enhanced safety and performance of nuclear power plants through improved service maintenance

    E-Print Network [OSTI]

    Achorn, Ted Glen

    1991-01-01

    Refinements in the safety and performance of nuclear power plants must be made to maintain public confidence and ensure competitiveness with other power sources. The aircraft industry, US Navy, and other programs have ...

  3. [6450-01-P], DEPARTMENT OF ENERGY, 10 CFR Part 830, Nuclear Safety Management, AGENCY: Department of Energy (DOE).

    Broader source: Energy.gov [DOE]

    The Department of Energy (DOE) is issuing a final rule regarding Nuclear SafetyManagement. This Part establishes requirements for the safe management of DOE contractor andsubcontractor work at the...

  4. Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository - BMT1 of the DECOVALEX III project. Part 3: Effects of THM coupling in sparsely fractured rocks

    E-Print Network [OSTI]

    2004-01-01

    Safety of a Hypothetical Nuclear Waste Repository – BMT1 ofsafety of a hypothetical nuclear waste repository – BMT1 ofsafety of a hypothetical nuclear waste repository – BMT1 of

  5. Locations of criticality alarms and nuclear accident dosimeters at Hanford. Revision 1

    SciTech Connect (OSTI)

    Not Available

    1992-08-01

    Hanford facilities that contain fissionable materials capable of achieving critical mass are monitored with nuclear accident dosimeters (NADS) in compliance with the requirements of DOE Order 5480.11, Chapter XI, Section 4.c. (DOE 1988). The US Department of Energy (DOE) Richland Field Office (RL) has assigned the responsibility for maintaining and evaluating the Hanford NAD system to the Instrumentation and External Dosimetry (I&ED) Section of Pacific Northwest Laboratory`s (PNL`s) Health Physics Department. This manual provides a description of the Hanford NAD, criteria and instructions for proper NAD placement, and the locations of these dosimeters onsite.

  6. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

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

  7. Formal Methods in Safety-Critical Railway Systems Thierry Lecomte 1

    E-Print Network [OSTI]

    Southampton, University of

    process and safety life cycle management. 2 French authorities define Qualified as « Certified and working an academic project). Event-B is the synthesis between B and Action System. Itt extends the usage of B of the outcome of Event-B is the proved definition of systems architectures and, more generally, the proved

  8. Professional Development Leave Report 1 Software as a Component in Safety-Critical Systems

    E-Print Network [OSTI]

    Vallino, James R

    1. Introduction The theme for my professional development leave was to consider software usage the software engineer must consider for standard desktop applications. In the real-time and embedded systems://www.se.rit.edu/~jrv/publications/Vallino-SoftwareSafety- LeaveProposal.pdf). The work was still directly related to the theme I proposed, but, for a variety

  9. Sandia Energy - Risk and Safety Assessment

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Risk and Safety Assessment Home Stationary Power Nuclear Fuel Cycle Nuclear Energy Safety Technologies Risk and Safety Assessment Risk and Safety AssessmentTara...

  10. International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6) Nara, Japan, October 4-8, 2004.

    E-Print Network [OSTI]

    control flow testing coverage criteria to the graph. With our approach, PLC based software designed in FBD coverage criteria to the graph. With 1 #12;The 6th International Conference on Nuclear Thermal HydraulicsThe 6th International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6

  11. Development of a Societal-Risk Goal for Nuclear Power Safety

    SciTech Connect (OSTI)

    Vicki Bier; Michael Corradini; Robert Youngblood; Caleb Roh; Shuji Liu

    2014-06-01

    The safety-goal policy of the Nuclear Regulatory Commission (NRC) has never included a true societal-risk goal. The NRC did acknowledge that the original goal for the risk of latent cancer facilities “was an individual risk goal not related to the number of people involved,” and stated that “a true societal risk goal would place a limit on the aggregate number of people affected.” However, this limitation was never satisfactorily addressed. Moreover, the safety goal has historically focused primarily on fatalities and latent health effects, while experience with actual nuclear accidents has shown that societal disruption can be significant even in accidents that yield only small to modest numbers of fatalities. Therefore, we have evaluated the social disruption effects from severe reactor accidents as a basis to develop a societal-risk goal for nuclear power plants, considering both health effects and non-health concerns such as property damage and land interdiction. Our initial analysis considered six different nuclear power plant sites in the U.S. for Boiling Water Reactors and Pressurized Water Reactors. The accident sequences considered for these two reactor types were station blackout sequences (both short-term and long-term SBO) as well as an STSBO with RCIC failure for the BWR and a Steam Generator Tube Rupture for the PWR. The source term release was an input in a RASCAL calculation of the off-site consequences using actual site-based weather data for each of the six plant sites randomly selected over a two-year period. The source term release plumes were then compared to Geographical Information System data for each site to determine the population affected and that would need to be evacuated to meet current emergency preparedness regulations. Our results to date suggest that number of people evacuated to meet current protective action guidelines appears to be a good proxy for disruption -- and, unlike other measures of disruption, has the advantage of being relatively straightforward to calculate for a given accident scenario and a given geographical location and plant site. Revised safety goals taking into account the potential for societal disruption could in principle be applied to the current generation of nuclear plants, but could also be used in evaluating and siting new technologies, such as small modular light water reactors, advanced Gen-IV high-temperature reactors, as well as reactor designs with passive safety features such as filtered vented containments.

  12. Annual report to Congress. Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, calendar year 2000

    SciTech Connect (OSTI)

    2001-03-01

    This Annual Report to the Congress describes the Department of Energy's activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board. During 2000, the Department completed its implementation and proposed closure of one Board recommendation and completed all implementation plan milestones associated with two additional Board recommendations. Also in 2000, the Department formally accepted two new Board recommendations and developed implementation plans in response to those recommendations. The Department also made significant progress with a number of broad-based safety initiatives. These include initial implementation of integrated safety management at field sites and within headquarters program offices, issuance of a nuclear safety rule, and continued progress on stabilizing excess nuclear materials to achieve significant risk reduction.

  13. Structural aging program to assess the adequacy of critical concrete components in nuclear power plants

    SciTech Connect (OSTI)

    Naus, D.J.; Marchbanks, M.F.; Oland, C.B.; Arndt, E.G.

    1989-01-01

    The Structural Aging (SAG) Program is carried out by the Oak Ridge National Laboratory (ORNL) under sponsorship of the United States Nuclear Regulatory Commission (USNRC). The Program has evolved from preliminary studies conducted to evaluate the long-term environmental challenges to light-water reactor safety-related concrete civil structures. An important conclusion of these studies was that a damage methodology, which can provide a quantitative measure of a concrete structure's durability with respect to potential future requirements, needs to be developed. Under the SAG Program, this issue is being addressed through: establishment of a structural materials information center, evaluation of structural component assessment and repair technologies, and development of a quantitative methodology for structural aging determinations. Progress to date of each of these activities is presented as well as future plans. 7 refs., 5 figs.

  14. Said Mughaghab Nuclear Data 2010,

    E-Print Network [OSTI]

    Danon, Yaron

    Said Mughaghab Nuclear Data 2010, Jeju Island, Korea, April 26-30, 2010 ANALYSIS OF MEASUREMENTS IN THE UNRESOLVED ENERGY REGION FOR ENDF EVALUATIONS RPI Nuclear Data 2011 Symposium for Criticality Safety and Reactor applications April 27, 2011 S. F. Mughabghab* National Nuclear Data Center Brookhaven National

  15. Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann Jackson About UsEnergy Marketing Corp. |Storage, Oversight Assessment of Nuclear Safety

  16. AREVA NP next generation fresh UO{sub 2} fuel assembly shipping cask: SCALE - CRISTAL comparisons lead to safety criticality confidence

    SciTech Connect (OSTI)

    Doucet, M.; Landrieu, M.; Montgomery, R.; O' Donnell, B.

    2007-07-01

    AREVA NP as a worldwide PWR fuel provider has to have a fleet of fresh UO{sub 2} shipping casks being agreed within a lot of countries including USA, France, Germany, Belgium, Sweden, China, and South Africa - and to accommodate foreseen EPR Nuclear Power Plants fuel buildings. To reach this target the AREVA NP Fuel Sector decided to develop an up-to-date shipping cask (so called MAP project) gathering experience feedback of the today fleet and an improved safety allowing the design to comply with international regulations (NRC and IAEA) and local Safety Authorities. Based on pre design features a safety case was set up to highlight safety margins. Criticality hypothetical accidental assumptions were defined: - Preferential flooding; - Fuel rod lattice pitch expansion for full length of fuel assemblies; - Neutron absorber penalty; -... Well known computer codes, American SCALE package and French CRISTAL package, were used to check configurations reactivity and to ensure that both codes lead to coherent results. Basic spectral calculations are based on similar algorithms with specific microscopic cross sections ENDF/BV for SCALE and JEF2.2 for CRISTAL. The main differences between the two packages is on one hand SCALE's three dimensional fuel assembly geometry is described by a pin by pin model while an homogenized fuel assembly description is used by CRISTAL and on the other hand SCALE is working with either 44 or 238 neutron energy groups while CRISTAL is with a 172 neutron energy groups. Those two computer packages rely on a wide validation process helping defining uncertainties as required by regulations in force. The shipping cask with two fuel assemblies is designed to maximize fuel isolation inside a cask and with neighboring ones even for large array configuration cases. Proven industrial products are used: - Boral{sup TM} as neutron absorber; - High density polyethylene (HDPE) or Nylon as neutron moderator; - Foam as thermal and mechanical protection. The cask is designed to handle the complete AREVA NP fuel assembly types from the 14x14 to the 18x18 design with a {sup 235}U enrichment up to 5.0% enriched natural uranium (ENU) and enriched reprocessed uranium (ERU). After a brief presentation of the computer codes and the description of the shipping cask, calculation results and comparisons between SCALE and CRISTAL will be discussed. (authors)

  17. Operation Cornerstone onsite radiological safety report for announced nuclear tests, October 1988--September 1989

    SciTech Connect (OSTI)

    Not Available

    1990-08-01

    Cornerstone was the name assigned to the series of underground nuclear experiments conducted at the Nevada Test Site (NTS) from October 1, 1988, through September 30, 1989. This report includes those experiments publicly announced. Remote radiation measurements were taken during and after each nuclear experiment by a telemetry system. Radiation Protection Technicians (RPT) with portable radiation detection instruments surveyed reentry routes into ground zeros (GZ) before other planned entries were made. Continuous surveillance was provided while personnel were in radiation areas and appropriate precautions were taken to protect persons from unnecessary exposure to radiation and toxic gases. Protective clothing and equipment were issued as needed. Complete radiological safety and industrial hygiene coverage were provided during drilling and mineback operations. Telemetered and portable radiation detector measurements are listed. Detection instrumentation used is described and specific operational procedures are defined.

  18. Risk-Informing Safety Reviews for Non-Reactor Nuclear Facilities

    SciTech Connect (OSTI)

    Mubayi, V.; Azarm, A.; Yue, M.; Mukaddam, W.; Good, G.; Gonzalez, F.; Bari, R.A.

    2011-03-13

    This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction help to risk-inform the safety review process and assess compliance with regulatory requirements.

  19. Use of artificial intelligence to enhance the safety of nuclear power plants

    SciTech Connect (OSTI)

    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.

  20. Safety

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    General Employee Radiological HS4240-W Chemical Safety HS4680-W PPE To access these training modules link here LTRAIN from inside LLNL, or here from anywhere. All JLF...

  1. John T. Mihalczo PhD. in Nuclear Engineering : University of Tennessee, 1970

    E-Print Network [OSTI]

    security by affecting safer assembly and handling of nuclear weapons components materials. The technologies in nuclear criticality safety, nuclear weapons identification, nuclear materials processing, and nuclearJohn T. Mihalczo PhD. in Nuclear Engineering : University of Tennessee, 1970 Masters in Physics

  2. MCNP6 Results for the Phase III Sensitivity Benchmark of the OCED/NEA Expert Group on Uncertainty Analysis for Criticality Safety Assessment

    SciTech Connect (OSTI)

    Kiedrowski, Brian C.

    2012-06-19

    Within the last decade, there has been increasing interest in the calculation of cross section sensitivity coefficients of k{sub eff} for integral experiment design and uncertainty analysis. The OECD/NEA has an Expert Group devoted to Sensitivity and Uncertainty Analysis within the Working Party for Nuclear Criticality Safety. This expert group has developed benchmarks to assess code capabilities and performance for doing sensitivity and uncertainty analysis. Phase III of a set of sensitivity benchmarks evaluates capabilities for computing sensitivity coefficients. MCNP6 has the capability to compute cross section sensitivities for k{sub eff} using continuous-energy physics. To help verify this capability, results for the Phase III benchmark cases are generated and submitted to the Expert Group for comparison. The Phase III benchmark has three cases: III.1, an array of MOX fuel pins, III.2, a series of infinite lattices of MOX fuel pins with varying pitches, and III.3 two spheres with homogeneous mixtures of UF{sub 4} and polyethylene with different enrichments.

  3. MCNP benchmark analyses of critical experiments for the Space Nuclear Thermal Propulsion program

    SciTech Connect (OSTI)

    Selcow, E.C.; Cerbone, R.J.; Ludewig, H.; Mughabghab, S.F.; Schmidt, E.; Todosow, M. [Brookhaven National Lab., Upton, NY (United States); Parma, E.J. [Sandia National Labs., Albuquerque, NM (United States); Ball, R.M.; Hoovler, G.S. [Babcock and Wilcox Co., Lynchburg, VA (United States)

    1993-06-01

    Benchmark analyses have been performed of Particle Bed Reactor (PBR) critical experiments (CX) using the MCNP radiation transport code. The experiments have been conducted at the Sandia National Laboratory reactor facility in support of the Space Nuclear Thermal Propulsion (SNTP) program. The test reactor is a nineteen element water moderated and reflected thermal system. A series of integral experiments have been carried out to test the capabilities of the radiation transport codes to predict the performance of PBR systems. MCNP was selected as the preferred radiation analysis tool for the benchmark experiments. Comparison between experimental and calculational results indicate very good agreement. This paper describes the analyses of benchmark experiments designed to quantify the accuracy of the MCNP radiation transport code for predicting the performance characteristics of PBR reactors.

  4. Updating Human Factors Engineering Guidelines for Conducting Safety Reviews of Nuclear Power Plants

    SciTech Connect (OSTI)

    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.

  5. Subject: Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities Project Number: 689Nuclear Energy Institute (NEI) Letter, 9/10/10

    Broader source: Energy.gov [DOE]

    Enclosed for your review is a Nuclear Energy Institute white paper on the use of Integrated Safety Analysis (ISA) at U.S. Nuclear Regulatory Commission-licensed recycling facilities. This paper is...

  6. Experimental study of neutron noise with criticality safety applications in mind

    SciTech Connect (OSTI)

    Barnett, C.S.

    1985-11-01

    A study has been conducted on the statistics of detected neutrons that leaked from four subcritical reflected, enriched-uranium assemblies, to explore the feasibility of developing a criticality warning system based on neutron noise analysis. Studies were conducted on three possible discriminators, i.e., three signatures that might be used to discriminate among assemblies of various multiplications. The noise analysis techniques studied performed well enough in deeply subcritical situations to deserve testing in an applications environment. They have a good chance of detecting changes in reactivity that are potentially dangerous. One can expect sharpest results when doing comparisons, i.e., when comparing two records, one taken in the past under circumstances known to be normal and one taken now to search for change.

  7. Quarterly report on Defense Nuclear Facilities Safety Board Recommendation 90-7 for the period ending December 31, 1992

    SciTech Connect (OSTI)

    Cash, R.J.; Dukelow, G.T.; Forbes, C.J.

    1993-03-01

    This is the seventh quarterly report on the progress of activities addressing safety issues associated with Hanford Site high-level radioactive waste tanks that contain ferrocyanide compounds. In the presence of oxidizing materials, such as nitrates or nitrites, ferrocyanide can be made to explode in the laboratory by heating it to high temperatures [above 285{degrees}C (545{degrees}F)]. In the mid 1950s approximately 140 metric tons of ferrocyanide were added to 24 underground high-level radioactive waste tanks. An implementation plan (Cash 1991) responding to the Defense Nuclear Facilities Safety Board Recommendation 90-7 (FR 1990) was issued in March 1991 describing the activities that were planned and underway to address each of the six parts of Recommendation 90-7. A revision to the original plan was transmitted to US Department of Energy by Westinghouse Hanford Company in December 1992. Milestones completed this quarter are described in this report. Contents of this report include: Introduction; Defense Nuclear Facilities Safety Board Implementation Plan Task Activities (Defense Nuclear Facilities Safety Board Recommendation for enhanced temperature measurement, Recommendation for continuous temperature monitoring, Recommendation for cover gas monitoring, Recommendation for ferrocyanide waste characterization, Recommendation for chemical reaction studies, and Recommendation for emergency response planning); Schedules; and References. All actions recommended by the Defense Nuclear Facilities Safety Board for emergency planning by Hanford Site emergency preparedness organizations have been completed.

  8. Annual report to Congress: Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, calendar year 1998

    SciTech Connect (OSTI)

    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.

  9. Annual report to Congress: Department of Energy activities relating to the Defense Nuclear Facilities Safety Board, Calendar Year 1999

    SciTech Connect (OSTI)

    None

    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.

  10. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01

    power plants must meet nuclear safety standards and adoptapplications; review of nuclear safety regula- tions; andpower development plans. Nuclear safety was placed front and

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

    SciTech Connect (OSTI)

    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 cost of decommissioning a PWR is lowest for ENTOMB and highest for SAFSTOR • the estimated cost of decommissioning a BWR is lowest for OECON and highest for SAFSTOR. In all cases, SAFSTOR has the lowest occupational radiation dose and the highest cost.

  12. Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization

    SciTech Connect (OSTI)

    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.

  13. Development of a Method for Quantifying the Reliability of Nuclear Safety-Related Software

    SciTech Connect (OSTI)

    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.

  14. Use of probabilistic risk assessment in expert system usage for nuclear power plant safety

    SciTech Connect (OSTI)

    Uhrig, R.E.

    1987-01-01

    The introduction of probability risk assessments (PRA's) to nuclear power plants in the Rasmussen Report (WASH-1400) gave us a means of evaluating the risk to the public associated with the operation of nuclear power plants, at least on a relative basis. While the choice of the ''source term'' and methodology in a PRA significantly influence the absolute probability and the consequences of core melt, comparison of two PRA calculations for two configurations of the same plant, carried out on a consistent basis, can be readily identify the increase in risk associated with going from one configuration of a plant to another by removing components or systems from service. This ratio of core melt probabilities (assuming no recovery of failed systems) obtained from two PRA calculations for different configurations was the criterion (called ''risk factor'') chosen as a basis for making a decision in an expert system as to what mitigating action, if any, would be taken to avoid a trip situation from developing. PRISIM was developed by JBF Associates of Knoxville under the sponsorship of the NRC as a system for Resident Inspectors at nuclear power plants to provide them with a relative safety status of the plant under all configurations. PRISIM calculated the risk factor---the ration of core melt probabilities of the plant under the current configuration relative to the normal configuration with all systems functioning---using an algorithm that emulates the results of the original PRA. It also presents time and core melt (assuming no recovery of systems or components).

  15. Validation of criticality safety calculational methods for U-AVLIS plant project

    SciTech Connect (OSTI)

    Lewis, K.D.

    1993-07-14

    The objectives of the Uranium Atomic Vapor Laser isotope Separation (U-AVLIS) are to develop, demonstrate, and deploy a laser-based process to enrich natural uranium in the U-235 isotope to levels useful as fuel in commercial light-water power reactors. Current U-AVLIS production plant criteria call for uranium product enriched in {sup 235}U up to 5 wt%. Development of the U-AVLIS technology is in an advanced stage, and demonstration of the integrated enrichment process is currently in progress using plant-scale equipment in the Uranium Demonstration System (UDS) at Lawrence Livermore National Laboratory. In this paper several existing experimental data which are applicable to the critical systems of importance to the safe design of the U-AVLIS plant are identified. These were used to benchmark a configuration-controlled, work station based version of one state-of-the-art computer code employed by the U-AVLIS program in UDS equipment design, and in U-AVLIS plant conceptual design NCS analyses.

  16. Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors

    SciTech Connect (OSTI)

    Biswas, D; Mennerdahl, D

    2008-06-23

    The ANSI/ANS 8.12 standard was first approved in July 1978. At that time, this edition was applicable to operations with plutonium-uranium oxide (MOX) fuel mixtures outside reactors and was limited to subcritical limits for homogeneous systems. The next major revision, ANSI/ANS-8.12-1987, included the addition of subcritical limits for heterogeneous systems. The standard was subsequently reaffirmed in February 1993. During late 1990s, substantial work was done by the ANS 8.12 Standard Working Group to re-examine the technical data presented in the standard using the latest codes and cross section sets. Calculations performed showed good agreement with the values published in the standard. This effort resulted in the reaffirmation of the standard in March 2002. The standard is currently in a maintenance mode. After 2002, activities included discussions to determine the future direction of the standard and to follow the MOX standard development by the International Standard Organization (ISO). In 2007, the Working Group decided to revise the standard to extend the areas of applicability by providing a wider range of subcritical data. The intent is to cover a wider domain of MOX fuel fabrication and operations. It was also decided to follow the ISO MOX standard specifications (related to MOX density and isotopics) and develop a new set of subcritical limits for homogeneous systems. This has resulted in the submittal (and subsequent approval) of the project initiation notification system form (PINS) in 2007.

  17. DOE-STD-1135-99 Guidance for Nuclear Criticality Safety Engineer Training and Qualification

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum Based| Department8, 2015 GATEWAY6.1viiiDepartmentMaster120-2005 Volume5-99

  18. Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Nonreactor Nuclear Facilities

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum12,ExecutiveFinancing Programs | Department ofIntegratingGuidelines

  19. Study of neutron noise from reflected, metal assemblies with criticality safety applications in mind

    SciTech Connect (OSTI)

    Barnett, C.S.

    1985-08-20

    The author studied the statistics of detected neutrons that leaked from four subcritical reflected, enriched-uranium assemblies, to explore the feasibility of developing a criticality warning system based on neutron noise analysis. The calculated multiplication factors of the assemblies are 0.59, 0.74, 0.82, and 0.92. The author studied three possible discriminators, i.e., three signatures that might be used to discriminate among assemblies of various multiplications. They are: (1) variance-to-mean ratio of the counts in a time bin (V/M); (2) covariance-to-mean ratio of the counts in a common time bin from two different detectors (C/M); and (3) covariance-to-mean ratio of the counts from a single detector in two adjacent time bins of equal length, which the author calls the serial-covariance-to-mean ratio (SC/M). The performances of the three discriminators were not greatly different, but a hierarchy did emerge: SC/M greater than or equal to V/M greater than or equal to C/M. An example of some results: in the neighborhood of k = 0.6 the ..delta..k required for satisfactory discrimination varies from about 3% to 7% as detector solid angle varies from 19% to 5%. In the neighborhood of k = 0.8 the corresponding ..delta..ks are 1% and 2%. The noise analysis techniques studied performed well enough in deeply subcritical situations to deserve testing in an applications environment. They have a good chance of detecting changes in reactivity that are potentially dangerous. One can expect sharpest results when doing comparisons, i.e., when comparing two records, one taken in the past under circumstances known to be normal and one taken now to search for change.

  20. Supporting Design of Safety-Critical Systems Dr Mark Nicholson, MATISSE Project (GR/R70590/01), University of York, UK

    E-Print Network [OSTI]

    Nicholson, Mark

    1, is likely to be a more appropriate development life cycle. Units Systems Platform ImplementationSupporting Design of Safety-Critical Systems Dr Mark Nicholson, MATISSE Project (GR/R70590 application areas, they may also lead to large economic losses, and even loss of human life. A computing

  1. Combining Total Monte Carlo and Benchmarks for nuclear data uncertainty propagation on an LFRs safety parameters

    E-Print Network [OSTI]

    Alhassan, Erwin; Duan, Junfeng; Gustavsson, Cecilia; Koning, Arjan; Pomp, Stephan; Rochman, Dimitri; Österlund, Michael

    2013-01-01

    Analyses are carried out to assess the impact of nuclear data uncertainties on keff for the European Lead Cooled Training Reactor (ELECTRA) using the Total Monte Carlo method. A large number of Pu-239 random ENDF-formated libraries generated using the TALYS based system were processed into ACE format with NJOY99.336 code and used as input into the Serpent Monte Carlo neutron transport code to obtain distribution in keff. The keff distribution obtained was compared with the latest major nuclear data libraries - JEFF-3.1.2, ENDF/B-VII.1 and JENDL-4.0. A method is proposed for the selection of benchmarks for specific applications using the Total Monte Carlo approach. Finally, an accept/reject criterion was investigated based on chi square values obtained using the Pu-239 Jezebel criticality benchmark. It was observed that nuclear data uncertainties in keff were reduced considerably from 748 to 443 pcm by applying a more rigid acceptance criteria for accepting random files.

  2. Combining Total Monte Carlo and Benchmarks for nuclear data uncertainty propagation on an LFRs safety parameters

    E-Print Network [OSTI]

    Erwin Alhassan; Henrik Sjöstrand; Junfeng Duan; Cecilia Gustavsson; Arjan Koning; Stephan Pomp; Dimitri Rochman; Michael Österlund

    2013-04-04

    Analyses are carried out to assess the impact of nuclear data uncertainties on keff for the European Lead Cooled Training Reactor (ELECTRA) using the Total Monte Carlo method. A large number of Pu-239 random ENDF-formated libraries generated using the TALYS based system were processed into ACE format with NJOY99.336 code and used as input into the Serpent Monte Carlo neutron transport code to obtain distribution in keff. The keff distribution obtained was compared with the latest major nuclear data libraries - JEFF-3.1.2, ENDF/B-VII.1 and JENDL-4.0. A method is proposed for the selection of benchmarks for specific applications using the Total Monte Carlo approach. Finally, an accept/reject criterion was investigated based on chi square values obtained using the Pu-239 Jezebel criticality benchmark. It was observed that nuclear data uncertainties in keff were reduced considerably from 748 to 443 pcm by applying a more rigid acceptance criteria for accepting random files.

  3. Quarterly report on Defense Nuclear Facilities Safety Board recommendation 90-7 for the period ending September 30, 1993

    SciTech Connect (OSTI)

    Meacham, J.E.; Cash, R.J.; Dukelow, G.T.

    1993-12-01

    This is the tenth quarterly repose on the progress of activities addressing safety issues associated with Hanford Site high-level radioactive waste tanks that contain ferrocyanide compounds. In the Presence of oxidizing materials, such as nitrates or nitrites, ferrocyanide can be made to explode in the laboratory by hearing it to high temperatures [above 285{degree} C (545{degree} F)]. In the mid 1950s, approximately 140 metric tons of ferrocyanide were added to waste now stored in underground high-level radioactive waste tanks. An implementation plan responding to the Defense Nuclear Facilities Safety Board Recommendation 90-7 (FR 1990){sup 2} was issued in March 1991 describing the activities that were planned and underway to address each of the six parts of Recommendation 90-7. A revision to the original plan was transmitted to the US Department of Energy by Westinghouse Hanford in December 1992, and subsequently to the Defense Nuclear Facilities Safety Board in 1993.

  4. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    Potential Safety Issues – Regulatory Design Criteria3-3 Regulatory design criteria for safety Table 3-4 Input3-4 Regulatory Design Criteria for safety The DRACS system

  5. Safety apparatus for nuclear reactor to prevent structural damage from overheating by core debris

    DOE Patents [OSTI]

    Gabor, John D. (Western Springs, IL); Cassulo, John C. (Stickney, IL); Pedersen, Dean R. (Naperville, IL); Baker, Jr., Louis (Downers Grove, IL)

    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.

  6. Environment, Safety and Health Self-Assessment Report Fiscal Year 2010

    E-Print Network [OSTI]

    Robinson, Scott

    2011-01-01

    Accreditation Program Nuclear Safety Management (Inventory)Exposure Nuclear Science Division Ergonomics Safety Physical

  7. Software Certification Experience in the Canadian Nuclear Industry: Lessons for the Future

    E-Print Network [OSTI]

    Lawford, Mark

    Keywords nuclear, software certification, safety-critical software 1. INTRODUCTION The computer controlled the first nuclear safety systems in Canada that were software based. This posed a problem for the regulatorsSoftware Certification Experience in the Canadian Nuclear Industry: Lessons for the Future Alan

  8. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    L. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,R. J. Neuhold, Introductury Nuclear Reactor Dynamics. ANSL. J. Hamilton Nuclear Reactor Analysis John Wiley and Sons,

  9. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    heat transfer. Nuclear Technology 163 (2008) 344- [18] V.and Electricity . Nuclear Technology 144 [5] D. F. Williams,Vessel Technology . Nuclear Technology, 78 (1987) 245- [5

  10. Systematic investigation of the effects of hydrophilic porosity on boiling heat transfer and critical heat flux

    E-Print Network [OSTI]

    Tetreault-Friend, Melanie

    2014-01-01

    Predicting the conditions of critical heat flux (CHF) is of considerable importance for safety and economic reasons in heat transfer units, such as in nuclear power plants. It is greatly advantageous to increase this thermal ...

  11. A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems

    E-Print Network [OSTI]

    Tree Analysis), FMEA (Failure Mode and Effect Analysis), HAZOP (Hazard and Operability study). · Safety

  12. Quarterly report on Defense Nuclear Facilities Safety Board Recommendation 90-7

    SciTech Connect (OSTI)

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

    1991-10-01

    This is the second quarterly report on the progress of activities addressing safety issues associated with Hanford Site high-level radioactive waste tanks that contain ferrocyanide compounds. An implementation plan (Cash 1991) responding to the Defense Nuclear Facilities Safety Board Recommendation 90-7 (FR 1990) was issued in March 1991 describing the activities planned and underway to address each of the six parts of the recommendation. All of the activities listed in the implementation plan are underway, including the multifunctional instrument tree and infrared tasks which resumed in late July. Although technical difficulties and resource limitations delayed some work, noteworthy progress has been made in completing a number of ferrocyanide program milestones. Thermal modeling shows that the heat loading of tank 241-BY-104 is much lower than previously listed and that significant hot spots within the waste are highly unlikely, if not possible. Computerized continuous temperature monitoring was installed on schedule on five of the highest interest tanks and five additional tanks will be on-line in December. Tank intrusive sampling is proceeding and the first vapor samples were obtained in mid-October. Spectral scans were completed for twelve tanks and substantial progress was made on design of an infrared scanning system. Chemical reaction studies are underway with synthetic ferrocyanide compounds believed to be more representative of the ferrocyanide materials actually deposited in the tanks. Tests indicate that water plays a major role in impeding a possible runaway reaction. An emergency planning exercise was conducted in May, emergency procedures were updated this quarter, and validation of the procedures and a second emergency exercise is scheduled for October.

  13. U.S. Department of Energy, Oak Ridge Operations Office Nuclear Facility Safety Basis Fundamentals Self-Study Guide [Fulfills ORO Safety Basis Competency 1, 2 (Part 1), or 7 (Part 1)

    Broader source: Energy.gov [DOE]

    "This self-study guide provides an overview of safety basis terminology, requirements, and activities that are applicable to DOE and Oak Ridge Operations Office (ORO) nuclear facilities on the Oak...

  14. A REVIEW OF LIGHT-WATER REACTOR SAFETY STUDIES. VOLUME 3 OF THE FINAL REPORT ON HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    of Michigan for the Atomic Power Development Association).light-water power plants now the Atomic Energy resides withU.S. Atomic Energy Commission, "The, Safety of Nuclear Power

  15. Proceedings of the US Nuclear Regulatory Commission twentieth water reactor safety information meeting; Volume 2, Severe accident research, Thermal hydraulics

    SciTech Connect (OSTI)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1993-03-01

    This three-volume report contains papers presented at the Twentieth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 21--23, 1992. The papers describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 10 different papers presented by researchersfrom CEC, China, Finland, France, Germany, Japan, Spain and Taiwan. Selected papers have been processed separately for inclusion in the Energy Science and Technology Database.

  16. Transactions of the Korean Nuclear Society Autumn Meeting Pyeongchang, Korea, October 30-31, 2014

    E-Print Network [OSTI]

    A Translator Verification Technique for FPGA Software Development in Nuclear Power Plants Jaeyeob Kim , Eui- critical system such as RPS (Reactor Protection System) in NPP (Nuclear Power Plants). Increasing in nuclear power plants, which is one of the most safety critical systems. While the designer develops

  17. 1Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory Nuclear Data Research at RPI

    E-Print Network [OSTI]

    Danon, Yaron

    1Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory Nuclear Data Research for Criticality Safety and Reactor Applications Rensselaer Polytechnic Institute, April 27, 2011 #12;2Mechanical.P Barry, Dr. R.C Block, B. Epping #12;3Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner

  18. Software Engineering for Safety: A Roadmap Robyn R. Lutz

    E-Print Network [OSTI]

    Lutz, Robyn R.

    ­ critical systems from nuclear and defense applications to domains as diverse as implantable medical devices analysis, (2) safety requirements specification and analysis, (3) designing for safety, (4) testing, (5, and (6) collaboration with related fields. The criteria used to choose the problems in section 3

  19. Spent Nuclear Fuel (SNF) project Integrated Safety Management System phase I and II Verification Review Plan

    SciTech Connect (OSTI)

    CARTER, R.P.

    1999-11-19

    The U.S. Department of Energy (DOE) commits to accomplishing its mission safely. To ensure this objective is met, DOE issued DOE P 450.4, Safety Management System Policy, and incorporated safety management into the DOE Acquisition Regulations ([DEAR] 48 CFR 970.5204-2 and 90.5204-78). Integrated Safety Management (ISM) requires contractors to integrate safety into management and work practices at all levels so that missions are achieved while protecting the public, the worker, and the environment. The contractor is required to describe the Integrated Safety Management System (ISMS) to be used to implement the safety performance objective.

  20. ReseaRch at the University of Maryland Nuclear Safety Research at the University of Maryland

    E-Print Network [OSTI]

    Hill, Wendell T.

    been a complicated rise and fall for nuclear technology. The proliferation of nuclear power plants and nuclear weapons was followed by controversial accidents and regulation. Today, nuclear power is considered that analyze the risks involved in the use of nuclear energy. Understanding and Using Radiation The ionizing

  1. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

    SciTech Connect (OSTI)

    Schultz, Peter Andrew

    2011-12-01

    The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

  2. Processes and Procedures for Application of CFD to Nuclear Reactor Safety Analysis

    SciTech Connect (OSTI)

    Richard W. Johnson; Richard R. Schultz; Patrick J. Roache; Ismail B. Celik; William D. Pointer; Yassin A. Hassan

    2006-09-01

    Traditionally, nuclear reactor safety analysis has been performed using systems analysis codes such as RELAP5, which was developed at the INL. However, goals established by the Generation IV program, especially the desire to increase efficiency, has lead to an increase in operating temperatures for the reactors. This increase pushes reactor materials to operate towards their upper temperature limits relative to structural integrity. Because there will be some finite variation of the power density in the reactor core, there will be a potential for local hot spots to occur in the reactor vessel. Hence, it has become apparent that detailed analysis will be required to ensure that local ‘hot spots’ do not exceed safety limits. It is generally accepted that computational fluid dynamics (CFD) codes are intrinsically capable of simulating fluid dynamics and heat transport locally because they are based on ‘first principles.’ Indeed, CFD analysis has reached a fairly mature level of development, including the commercial level. However, CFD experts are aware that even though commercial codes are capable of simulating local fluid and thermal physics, great care must be taken in their application to avoid errors caused by such things as inappropriate grid meshing, low-order discretization schemes, lack of iterative convergence and inaccurate time-stepping. Just as important is the choice of a turbulence model for turbulent flow simulation. Turbulence models model the effects of turbulent transport of mass, momentum and energy, but are not necessarily applicable for wide ranges of flow types. Therefore, there is a well-recognized need to establish practices and procedures for the proper application of CFD to simulate flow physics accurately and establish the level of uncertainty of such computations. The present document represents contributions of CFD experts on what the basic practices, procedures and guidelines should be to aid CFD analysts to obtain accurate estimates of the flow and energy transport as applied to nuclear reactor safety. However, it is expected that these practices and procedures will require updating from time to time as research and development affect them or replace them with better procedures. The practices and procedures are categorized into five groups. These are: 1.Code Verification 2.Code and Calculation Documentation 3.Reduction of Numerical Error 4.Quantification of Numerical Uncertainty (Calculation Verification) 5.Calculation Validation. These five categories have been identified from procedures currently required of CFD simulations such as those required for publication of a paper in the ASME Journal of Fluids Engineering and from the literature such as Roache [1998]. Code verification refers to the demonstration that the equations of fluid and energy transport have been correctly coded in the CFD code. Code and calculation documentation simply means that the equations and their discretizations, etc., and boundary and initial conditions used to pose the fluid flow problem are fully described in available documentation. Reduction of numerical error refers to practices and procedures to lower numerical errors to negligible or very low levels as is reasonably possible (such as avoiding use of first-order discretizations). The quantification of numerical uncertainty is also known as calculation verification. This means that estimates are made of numerical error to allow the characterization of the numerical

  3. Nuclear safety analyses and core design calculations to convert the Texas A & M University Nuclear Science Center reactor to low enrichment uranium fuel. Final report

    SciTech Connect (OSTI)

    Parish, T.A.

    1995-03-02

    This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.

  4. Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process

    Broader source: Energy.gov [DOE]

    Presenter: Undine Shoop, Chief, Health Physics and Human Performance Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission

  5. Facts and Lessons of the Fukushima Nuclear Accident and Safety Improvement- The Operator Viewpoints

    Broader source: Energy.gov [DOE]

    Presenter: Akira Kawano, General Manager, Nuclear International Relations and Strategy Group, Nuclear Power and Plant Siting Administrative Department, Tokyo Electric Power Company

  6. Risk-informed incident management for nuclear power plants

    E-Print Network [OSTI]

    Smith, Curtis Lee, 1966-

    2002-01-01

    Decision making as a part of nuclear power plant operations is a critical, but common, task. Plant management is forced to make decisions that may have safety and economic consequences. Formal decision theory offers the ...

  7. Cognitive decision errors and organization vulnerabilities in nuclear power plant safety management: Modeling using the TOGA meta-theory framework

    SciTech Connect (OSTI)

    Cappelli, M.; Gadomski, A. M.; Sepiellis, M.; Wronikowska, M. W.

    2012-07-01

    In the field of nuclear power plant (NPP) safety modeling, the perception of the role of socio-cognitive engineering (SCE) is continuously increasing. Today, the focus is especially on the identification of human and organization decisional errors caused by operators and managers under high-risk conditions, as evident by analyzing reports on nuclear incidents occurred in the past. At present, the engineering and social safety requirements need to enlarge their domain of interest in such a way to include all possible losses generating events that could be the consequences of an abnormal state of a NPP. Socio-cognitive modeling of Integrated Nuclear Safety Management (INSM) using the TOGA meta-theory has been discussed during the ICCAP 2011 Conference. In this paper, more detailed aspects of the cognitive decision-making and its possible human errors and organizational vulnerability are presented. The formal TOGA-based network model for cognitive decision-making enables to indicate and analyze nodes and arcs in which plant operators and managers errors may appear. The TOGA's multi-level IPK (Information, Preferences, Knowledge) model of abstract intelligent agents (AIAs) is applied. In the NPP context, super-safety approach is also discussed, by taking under consideration unexpected events and managing them from a systemic perspective. As the nature of human errors depends on the specific properties of the decision-maker and the decisional context of operation, a classification of decision-making using IPK is suggested. Several types of initial situations of decision-making useful for the diagnosis of NPP operators and managers errors are considered. The developed models can be used as a basis for applications to NPP educational or engineering simulators to be used for training the NPP executive staff. (authors)

  8. Safety research programs sponsored by Office of Nuclear Regulatory Research. Quarterly progress report, July 1-September 30, 1983. Volume 3, No. 3

    SciTech Connect (OSTI)

    Weiss, A J [comp.

    1984-01-01

    The projects reported are the following: HTGR Safety Evaluation, SSC Development, Validation and Application, CRBR Balance of Plant Modeling, Thermal-Hydraulic Reactor Safety Experiments, LWR Plant Analyzer Development, LWR Code Assessment and Application, Thermal Reactor Code Development (RAMONA-3B); Stress Corrosion Cracking of PWR Steam Generator Tubing, Bolting Failure Analysis, Probability Based Load Combinations for Design of Category I Structures, Mechanical Piping Benchmark Problems; Human Error Data for Nuclear Power Plant Safety-Related Events, and Human Factors in Nuclear Power Plant Safeguards.

  9. Policy of EDF for the future of nuclear power generation safety and economy

    SciTech Connect (OSTI)

    Roche, B.

    1997-12-01

    EDF improves at the same time economy and safety of its existing units. For new designs, it is the same fight.

  10. Regulatory analysis for the resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants

    SciTech Connect (OSTI)

    Chang, T.Y.

    1991-09-01

    Generic Safety Issue (GSI)-29 deals with staff concerns about public risk due to degradation or failure of safety-related bolting in nuclear power plants. The issue was initiated in November 1982. Value-impact studies of a mandatory program on safety-related bolting for operating plants were inconclusive: therefore, additional regulatory requirements for operating plants could not be justified in accordance with provisions of 10 CFR 50.109. In addition, based on operating experience with bolting in both nuclear and conventional power plants, the actions already taken through bulletins, generic letters, and information notices, and the industry-proposed actions, the staff concluded that a sufficient technical basis exists for the resolution of GSI-29. The staff further concluded that leakage of bolted pressure joints is possible but catastrophic failure of a reactor coolant pressure boundary joint that will lead to significant accident sequences is highly unlikely. For future plants, it was concluded that a new Standard Review Plant section should be developed to codify existing bolting requirements and industry-developed initiatives. 9 refs., 1 tab.

  11. Review of nuclear power plant safety cable aging studies with recommendations for improved approaches and for future work.

    SciTech Connect (OSTI)

    Gillen, Kenneth Todd; Bernstein, Robert

    2010-11-01

    Many U. S. nuclear power plants are approaching 40 years of age and there is a desire to extend their life for up to 100 total years. Safety-related cables were originally qualified for nuclear power plant applications based on IEEE Standards that were published in 1974. The qualifications involved procedures to simulate 40 years of life under ambient power plant aging conditions followed by simulated loss of coolant accident (LOCA). Over the past 35 years or so, substantial efforts were devoted to determining whether the aging assumptions allowed by the original IEEE Standards could be improved upon. These studies led to better accelerated aging methods so that more confident 40-year lifetime predictions became available. Since there is now a desire to potentially extend the life of nuclear power plants way beyond the original 40 year life, there is an interest in reviewing and critiquing the current state-of-the-art in simulating cable aging. These are two of the goals of this report where the discussion is concentrated on the progress made over the past 15 years or so and highlights the most thorough and careful published studies. An additional goal of the report is to suggest work that might prove helpful in answering some of the questions and dealing with some of the issues that still remain with respect to simulating the aging and predicting the lifetimes of safety-related cable materials.

  12. Safety Evaluation Report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 12

    SciTech Connect (OSTI)

    Tam, P.S.

    1993-10-01

    Supplement No. 12 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation of (1) additional information submitted by the applicant since Supplement No. 11 was issued, and (2) matters that the staff had under review when Supplement No. 11 was issued.

  13. Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 14

    SciTech Connect (OSTI)

    Tam, P.S.

    1994-12-01

    Supplement No. 14 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation with additional information submitted by the applicant since Supplement No. 13 was issued, and matters that the staff had under review when Supplement No. 13 was issued.

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

    SciTech Connect (OSTI)

    Marshall, A.C.

    1995-12-31

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

  15. Quarterly report on Defense Nuclear Facilities Safety Board Recommendation 90-7 for the period ending March 31, 1993

    SciTech Connect (OSTI)

    Cash, R.J.; Dukelow, G.T.; Forbes, C.J.; Meacham, J.E.

    1993-06-01

    This is the eighth quarterly report on the progress of activities addressing safety issues associated with Hanford Site high-level radioactive waste tanks that contain ferrocyanide compounds. In the presence of oxidizing materials, such as nitrates or nitrites, ferrocyanide can be made to explode in the laboratory by heating it to high temperatures [above 285{degree}C (545{degree}F)]. In the mid 1950s, approximately 140 metric tons of ferrocyanide were added to 24 underground high-level radioactive waste tanks. An implementation plan (Cash 1991) responding to the Defense Nuclear Facilities Safety Board Recommendation 90-7 (FR 1990) was issued in March 1991 describing the activities that were planned and underway to address each of the six parts of Recommendation 90-7. A revision to the original plan was transmitted to the US Department of Energy by Westinghouse Hanford Company in December 1992.

  16. Quarterly report on Defense Nuclear Facilities Safety Board Recommendation 90-7 for the period ending March 31, 1992

    SciTech Connect (OSTI)

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

    1992-05-01

    This is the fourth quarterly report on the progress of activities addressing safety issues associated with Hanford Site high-level radioactive waste tanks that contain ferrocyanide compounds. In the presence of oxidizing materials, such as nitrates or nitrites, ferrocyanide can be made to explode in the laboratory by heating it to high temperatures (above 285{degrees}C (545{degree}F)). In the mid 1950s approximately 140 metric tons of ferrocyanide were added to 24 underground high-level radioactive waste tanks. An implementation plan (Cash 1991) responding to the Defense Nuclear Facilities Safety Board Recommendation 90-7 (FR 1990) was issued in March 1991 describing the activities planned and underway to address each of the six parts of the recommendation.

  17. Quarterly report on Defense Nuclear Facilities Safety Board Recommendation 90-7 for the period ending March 31, 1992

    SciTech Connect (OSTI)

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

    1992-05-01

    This is the fourth quarterly report on the progress of activities addressing safety issues associated with Hanford Site high-level radioactive waste tanks that contain ferrocyanide compounds. In the presence of oxidizing materials, such as nitrates or nitrites, ferrocyanide can be made to explode in the laboratory by heating it to high temperatures [above 285{degrees}C (545{degree}F)]. In the mid 1950s approximately 140 metric tons of ferrocyanide were added to 24 underground high-level radioactive waste tanks. An implementation plan (Cash 1991) responding to the Defense Nuclear Facilities Safety Board Recommendation 90-7 (FR 1990) was issued in March 1991 describing the activities planned and underway to address each of the six parts of the recommendation.

  18. An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

    SciTech Connect (OSTI)

    P. Calderoni; P. Sharpe; M. Shimada

    2009-09-01

    The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

  19. Evaluation of station blackout accidents at nuclear power plants: Technical findings related to unresolved safety issue A-44: Final report

    SciTech Connect (OSTI)

    Not Available

    1988-06-01

    ''Station Blackout,'' which is the complete loss of alternating current (AC) electrical power in a nuclear power plant, has been designated as Unresolved Safety Issue A-44. Because many safety systems required for reactor core decay heat removal and containment heat removal depend on AC power, the consequences of a station blackout could be severe. This report documents the findings of technical studies performed as part of the program to resolve this issue. The important factors analyzed include: the fequency of loss of offsite power; the probability that emergency or onsite AC power supplies would be unavailable; the capability and reliability of decay heat removal systems independent of AC power; and the likelihood that offsite power would be restored before systems that cannot operate for extended periods without AC power fail, thus resulting in core damage. This report also addresses effects of different designs, locations, and operational features on the estimated frequency of core damage resulting from station blackout events.

  20. Valuing the greenhouse gas emissions from nuclear power: A critical survey Benjamin K. Sovacool

    E-Print Network [OSTI]

    Laughlin, Robert B.

    estimates. It calculates that while the range of emissions for nuclear energy over the lifetime of a plant responsible for the disparity in lifecycle estimates, in particular identifying errors in both the lowest estimates (not comprehensive) and the highest estimates (failure to consider co-products). It should

  1. Nuclear Engineer

    Broader source: Energy.gov [DOE]

    This position is located in the Nuclear Safety Division (NSD) which has specific responsibility for managing the development, analysis, review, and approval of non-reactor nuclear facility safety...

  2. A review of the current state-of-the-art methodology for handling bias and uncertainty in performing criticality safety evaluations. Final report

    SciTech Connect (OSTI)

    Disney, R.K.

    1994-10-01

    The methodology for handling bias and uncertainty when calculational methods are used in criticality safety evaluations (CSE`s) is a rapidly evolving technology. The changes in the methodology are driven by a number of factors. One factor responsible for changes in the methodology for handling bias and uncertainty in CSE`s within the overview of the US Department of Energy (DOE) is a shift in the overview function from a ``site`` perception to a more uniform or ``national`` perception. Other causes for change or improvement in the methodology for handling calculational bias and uncertainty are; (1) an increased demand for benchmark criticals data to expand the area (range) of applicability of existing data, (2) a demand for new data to supplement existing benchmark criticals data, (3) the increased reliance on (or need for) computational benchmarks which supplement (or replace) experimental measurements in critical assemblies, and (4) an increased demand for benchmark data applicable to the expanded range of conditions and configurations encountered in DOE site restoration and remediation.

  3. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01

    safety limits for fuel and metallic structures in contactfuel temperature profile along the core (left) and DHX transient average metallicfuel temperature profile along the core (left) and DHX transient average metallic

  4. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2002-05-20

    To establish facility safety requirements for the Department of Energy, including National Nuclear Security Administration. Cancels DOE O 420.1. Canceled by DOE O 420.1B.

  5. PRECLOSURE CRITICALITY ANALYSIS PROCESS REPORT

    SciTech Connect (OSTI)

    A.E. Danise

    2004-10-25

    This report describes a process for performing preclosure criticality analyses for a repository at Yucca Mountain, Nevada. These analyses will be performed from the time of receipt of fissile material until permanent closure of the repository (preclosure period). The process describes how criticality safety analyses will be performed for various configurations of waste in or out of waste packages that could occur during preclosure as a result of normal operations or event sequences. The criticality safety analysis considers those event sequences resulting in unanticipated moderation, loss of neutron absorber, geometric changes, or administrative errors in waste form placement (loading) of the waste package. The report proposes a criticality analyses process for preclosure to allow a consistent transition from preclosure to postclosure, thereby possibly reducing potential cost increases and delays in licensing of Yucca Mountain. The proposed approach provides the advantage of using a parallel regulatory framework for evaluation of preclosure and postclosure performance and is consistent with the U.S. Nuclear Regulatory Commission's approach of supporting risk-informed, performance-based regulation for fuel cycle facilities, ''Yucca Mountain Review Plan, Final Report'', and 10 CFR Part 63. The criticality-related criteria for ensuring subcriticality are also described as well as which guidance documents will be utilized. Preclosure operations and facilities have significant similarities to existing facilities and operations currently regulated by the U.S. Nuclear Regulatory Commission; therefore, the design approach for preclosure criticality safety will be dictated by existing regulatory requirements while using a risk-informed approach with burnup credit for in-package operations.

  6. Levitation is a useful scientific tool to study materials in extreme, metastable and often tran-sient states, encompassing a wide range of disciplines from pharmaceutics to nuclear safety.

    E-Print Network [OSTI]

    Kilian, Kristopher A.

    - sient states, encompassing a wide range of disciplines from pharmaceutics to nuclear safety. Acoustic, simulating the melting of a nuclear fuel rod. Future plans to alter the oxidation state of levitat- ed oxide

  7. Analysis of muon radiography of the Toshiba nuclear critical assembly reactor

    SciTech Connect (OSTI)

    Morris, C. L.; Bacon, Jeffery; Borozdin, Konstantin; Fabritius, J. M.; Perry, John; Ramsey, John [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Ban, Yuichiro; Izumi, Mikio; Sano, Yuji; Yoshida, Noriyuki [Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Miyadera, Haruo [Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Toshiba Corporation, 8 Shinsugita-cho, Isogo-ku, Yokohama 235-8523 (Japan); Mizokami, Shinya; Otsuka, Yasuyuki; Yamada, Daichi [Tokyo Electric Power Company, 1-1-3 Uchisaiwai-cho, Chiyoda-ku, Tokyo (Japan); Sugita, Tsukasa; Yoshioka, Kenichi [Toshiba Corporation, 4-1 Ukishima-cho, Kawasaki-ku, Kawasaki 210-0862 (Japan)

    2014-01-13

    A 1.2?×?1.2 m{sup 2} muon tracker was moved from Los Alamos to the Toshiba facility at Kawasaki, Japan, where it was used to take ?4 weeks of data radiographing the Toshiba Critical Assembly Reactor with cosmic ray muons. In this paper, we describe the analysis procedure, show results of this experiment, and compare the results to Monte Carlo predictions. The results validate the concept of using cosmic rays to image the damaged cores of the Fukushima Daiichi reactors.

  8. National Criticality Experiments Research Center: Capability and Status

    SciTech Connect (OSTI)

    Hayes, David K. [Los Alamos National Laboratory; Myers, William L. [Los Alamos National Laboratory

    2012-07-12

    After seven years, the former Los Alamos Critical Experiments Facility (LACEF), or Pajarito Site, has reopened for business as the National Criticality Experiments Research Center (NCERC) at the Nevada National Security Site (NNSS). Four critical assembly machines (Comet, Planet, Flat-Top, and Godiva-IV) made the journey from Los Alamos to the NNSS. All four machines received safety system upgrades along with new digital control systems. Between these machines, systems ranging from the thermal through the intermediate to the fast spectrum may be assembled. Steady-State, transient, and super-prompt critical conditions may be explored. NCERC is the sole remaining facility in the United States capable of conducting general-purpose nuclear materials handling including the construction and operation of high-multiplication assemblies, delayed critical assemblies, and prompt critical assemblies. Reconstitution of the unique capabilities at NCERC ensures the viability of (1) The Nuclear Renaissance, (2) Stockpile Stewardship, and (3) and the next generation of criticality experimentalists.

  9. CRAD, Assessment Criteria and Guidelines for Determining the Adequacy of Software Used in the Safety Analysis and Design of Defense Nuclear Facilities

    Broader source: Energy.gov [DOE]

    These guidelines and criteria provide a consistent overall framework for assessment of the processes that are currently in place to ensure that the software being used in the safety analysis and design of the SSCs in defense nuclear facilities is adequate. These reviews will be conducted only on software that is currently in use, not on software that was previously used as part of a safety analysis and design process.

  10. Development of an Updated Societal-Risk Goal for Nuclear Power Safety

    SciTech Connect (OSTI)

    Vicki Bier; Michael Corradini; Robert Youngblood; Caleb Roh; Shuji Liu

    2014-07-01

    This report briefly summarizes work done in FY 2013 on the subject LDRD. The working hypothesis is that societal disruption should be addressed in a safety goal. This is motivated by the point that the Fukushima disaster resulted in very little public dose, but enormous societal disruption; a goal that addressed societal disruption would fill a perceived gap in the US NRC safety goal structure. This year's work entailed analyzing the consequences of postulated accidents at various reactor sites in the US, specifically with a view to quantifying the number of people relocated and the duration of their relocation, to see whether this makes sense as a measure of societal disruption.

  11. Guidelines for nuclear-power-plant safety-issue-prioritization information development

    SciTech Connect (OSTI)

    Andrews, W.B.; Gallucci, R.H.V.; Heaberlin, S.W.; Bickford, W.E.; Konzek, G.J.; Strenge, D.L.; Smith, R.I.; Weakley, S.A.

    1983-02-01

    Pacific Northwest Laboratory has developed a methodology, with examples, to calculate - to an approximation serviceable for prioritization purposes - the risk, dose and cost impacts of implementing resolutions to reactor safety issues. This report is an applications guide to issue-specific calculations. A description of the approach, mathematical models, worksheets and step-by-step examples are provided. Analysis using this method is intended to provide comparable results for many issues at a cost of two staff-weeks per issue. Results will be used by the NRC to support decisions related to issue priorities in allocation of resources to complete safety issue resolutions.

  12. Guidelines for nuclear power plant safety issue prioritization information development. Supplement 5

    SciTech Connect (OSTI)

    Daling, P.M.; Lavender, J.C.

    1996-07-01

    This is the sixth in a series of reports to document the development and use of a methodology developed by the Pacific Northwest Laboratory (PNL) to calculate, for prioritization purposes, the risk, dose, and cost impacts of implementing potential resolutions to reactor safety issues (see NUREG/CR-2800, Andrews, et al., 1983). This report contains the results of issue-specific analyses for 34 generic issues. Each issue was considered within the constraints of available information at the time the issues were examined and approximately 2 staff-weeks of labor. The results are referenced as one consideration in NUREG-0933, A Prioritization of Generic Safety Issues (Emrit, et al., 1983).

  13. Safety research programs sponsored by Office of Nuclear Regulatory Research. Quarterly progress report, October 1-December 31, 1983. Volume 3, No. 4

    SciTech Connect (OSTI)

    Weiss, A J [comp.

    1984-05-01

    The projects reported are the following: High Temperature Reactor Research, SSC Development, Validation and Application, CRBR Balance of Plant Modeling, Thermal-Hydraulic Reactor Safety Experiments, Development of Plant Analyzer, Code Assessment and Application (Transient and LOCA Analyses), Thermal Reactor Code Development (RAMONA-3B), Calculational Quality Assurance in Support of PTS; Stress Corrosion Cracking of PWR Steam Generator Tubing, Bolting Failure Analysis, Probability Based Load Combinations for Design of Category I Structures, Mechanical Piping Benchmark Problems, Identification of Age-Related Failure Modes; Analysis of Human Error Data for Nuclear Power Plant Safety-Related Events, Human Factors in Nuclear Power Plant Safeguards, Emergency Action Levels, and Protective Action Decision Making.

  14. 2013 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  15. 2014 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  16. 2011 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    Managers perform an annual workforce analysis of their organization and develop staffing plans that identify technical capabilities and positions they need to ensure safe operation of defense nuclear facilities.

  17. Safety Software Quality Assurance Functions, Responsibilities, and Authorities for Nuclear Facilities and Activities

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2003-08-27

    To assign roles and responsibilities for improving the quality of safety software. DOE N 411.2 (archived) extends this Notice until 01/31/2005. DOE N 411.3 extends this Notice until 1/31/06. Canceled by DOE O 414.1C. does not cancel other directives.

  18. AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

    SciTech Connect (OSTI)

    Gorgemans, J.; Mulhollem, L.; Glavin, J.; Pfister, A.; Conway, L.; Schulz, T.; Oriani, L.; Cummins, E.; Winters, J. [Westinghouse Electric Company LLC, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all possible pool heat load conditions. - After 3 days, several different means are provided to continue spent fuel cooling using installed plant equipment as well as off-site equipment with built-in connections. Even for beyond design basis accidents with postulated pool damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 multiple spent fuel pool spray and fill systems provide additional lines of defense to prevent spent fuel damage. (authors)

  19. The Critical Mass Laboratory at Rocky Flats

    SciTech Connect (OSTI)

    Rothe, Robert E

    2003-10-15

    The Critical Mass Laboratory (CML) at Rocky Flats northwest of Denver, Colorado, was built in 1964 and commissioned to conduct nuclear experiments on January 28, 1965. It was built to attain more accurate and precise experimental data to ensure nuclear criticality safety at the plant than were previously possible. Prior to its construction, safety data were obtained from long extrapolations of subcritical data (called in situ experiments), calculated parameters from reactor engineering 'models', and a few other imprecise methods. About 1700 critical and critical-approach experiments involving several chemical forms of enriched uranium and plutonium were performed between then and 1988. These experiments included single units and arrays of fissile materials, reflected and 'bare' systems, and configurations with various degrees of moderation, as well as some containing strong neutron absorbers. In 1989, a raid by the Federal Bureau of Investigation (FBI) caused the plant as a whole to focus on 'resumption' instead of further criticality safety experiments. Though either not recognized or not admitted for a few years, that FBI raid did sound the death knell for the CML. The plant's optimistic goal of resumption evolved to one of deactivation, decommissioning, and plantwide demolition during the 1990s. The once-proud CML facility was finally demolished in April of 2002.

  20. A primer for criticality calculations with DANTSYS

    SciTech Connect (OSTI)

    Busch, R.D.

    1997-08-01

    With the closure of many experimental facilities, the nuclear safety analyst has to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. Although deterministic methods often do not provide exact models of a system, a substantial amount of reliable information on nuclear systems can be obtained using these methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico (UNM) in cooperation with the Radiation Transport Group at Los Alamos National Laboratory (LANL) has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. DANTSYS is the new name of the group of codes formerly known as: ONEDANT, TWODANT, TWOHEX, TWOGQ, and THREEDANT. The primer is designed to teach bu example, with each example illustrating two or three DANTSYS features useful in criticality analyses. Starting with a Quickstart chapter, the primer gives an overview of the basic requirements for DANTSYS input and allows the user to quickly run a simple criticality problem with DANTSYS. Each chapter has a list of basic objectives at the beginning identifying the goal of the chapter and the individual DANTSYS features covered in detail in the chapter example problems. On completion of the primer, it is expected that the user will be comfortable doing criticality calculations with DANTSYS and can handle 60--80% of the situations that normally arise in a facility. The primary provides a set of input files that can be selective modified by the user to fit each particular problem.

  1. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01

    1986, 10 CFR Part 50. ). Defence in Depth in Nuclear Safety.Report by the International Nuclear Safety Advisory Group. (Ministerial Conference on Nuclear Safety. The Accident at

  2. 1Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory RPI LINAC Facility

    E-Print Network [OSTI]

    Danon, Yaron

    1Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory RPI LINAC Facility Nuclear Criticality Safety Program Conference April 27, 2011 #12;2Mechanical, Aerospace and Nuclear · Refurbishment/Upgrade Projects (Recent Future) · Resources at the Facility #12;3Mechanical, Aerospace

  3. Transactions of the Korean Nuclear Society Spring Meeting Jeju, Korea, May 17-18, 2012

    E-Print Network [OSTI]

    DE: Development Environment for Safety-Critical Software of Nuclear Power Plant Jong-Hoon Lee , Junbeom Yoo V&V [2], V&V tasks are defined for each development phase. Software in Nuclear Power Plant (NPPTransactions of the Korean Nuclear Society Spring Meeting Jeju, Korea, May 17-18, 2012 Nu

  4. Nuclear Fission Reactor Safety Research in FP7 and future perspectives

    E-Print Network [OSTI]

    Garbil, Roger

    2014-01-01

    The European Union (?U) has defined in the Europe 2020 strategy and 2050 Energy Roadmap its long-term vision for establishing a secure, sustainable and competitive energy system and setting up legally binding targets by 2020 for reducing greenhouse emissions, by increasing energy efficiency and the share of renewable energy sources while including a significant share from nuclear fission. Nuclear energy can enable the further reduction in harmful emissions and can contribute to the EU’s competitive energy system, security of supply and independence from fossil fuels. Nuclear fission is a valuable option for those 14 EU countries that promote its use as part of their national energy mix. The European Group on Ethics in Science and New Technologies (EGE) adopted its Opinion No.27 ‘An ethical framework for assessing research, production and use of energy’ and proposed an integrated ethics approach for the research, production and use of energy in the EU, seeking equilibrium among four criteria – access ...

  5. Independent criticality safety evaluation of deposits in cooler equipment in Building K-31 at the Oak Ridge K-25 Site, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    NONE

    1996-10-01

    This report provides an independent assessment of nuclear criticality issues associated with uranium deposits in the West and East Coolers for the 6A Booster Station in Building K-31 at the Oak Ridge K-25 Site. This assessment investigates the applicability of the initial assumptions used by Lockheed Martin Energy Systems (Energy Systems) and evaluates criticality calculations previously completed by Energy Systems. The calculations were independently verified. Each component was evaluated for its ability to satisfy requirements for subcriticality and meet the double contingency principle. Facility walk downs, detailed neutronics analysis, and fault tree analysis (FTA) were performed. The facility walk downs provided a better understanding of the building condition and status, equipment configuration, and uranium deposit locations. The detailed neutronics analysis focused on system geometry and moderation levels applicable to the individual components. The FTA considered the annual rate of occurrence for the events identified as potential causes of criticality issues. This report also examines the advantages of using this type of evaluation to assess the removal process for additional components and equipment.

  6. Safety Criteria and Safety Lifecycle for Artificial Neural Networks

    E-Print Network [OSTI]

    Kelly, Tim

    Safety Criteria and Safety Lifecycle for Artificial Neural Networks Zeshan Kurd, Tim Kelly and Jim performance based techniques that aim to improve the safety of neural networks for safety critical for safety assurance. As a result, neural networks are typically restricted to advisory roles in safety

  7. Aging and service wear of air-operated valves used in safety-related systems at nuclear power plants

    SciTech Connect (OSTI)

    Cox, D.F.; McElhaney, K.L.; Staunton, R.H.

    1995-05-01

    Air-operated valves (AOVs) are used in a variety of safety-related applications at nuclear power plants. They are often used where rapid stroke times are required or precise control of the valve obturator is required. They can be designed to operate automatically upon loss of power, which is often desirable when selecting components for response to design basis conditions. The purpose of this report is to examine the reported failures of AOVs and determine whether there are identifiable trends in the failures related to predictable causes. This report examines the specific components that comprise a typical AOV, how those components fail, when they fail, and how such failures are discovered. It also examines whether current testing frequencies and methods are effective in predicting such failures.

  8. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Measurements - Nuclear Regulatory Commission - protectiveand by the Nuclear Regulatory Commission (NRC) as a basisplants. The Nuclear Regulatory Commission is the agency

  9. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    DESIGN, TESTING, AND MAINTENANCE CRITERIA FOR ENGINEERED-SAFETY-design and operating criteria and to show the adequacy of the site characteristics from a safety

  10. International Symposium on Fusion Nuclear Technology (ISFNT-5) SAFETY ISSUES ASSOCIATED WITH MOBILIZED ACTIVATION

    E-Print Network [OSTI]

    California at Los Angeles, University of

    of the tin component. Screening Criteria The designs under development in the APEX project are at a pre project, we have developed a set of screening criteria that are used at early stages of design to identify Symposium on Fusion Nuclear Technology (ISFNT-5) A design must adequately transfer heat from plasma

  11. Preliminary systems-interaction results from the Digraph Matrix Analysis of the Watts Bar Nuclear Power Plant safety-injection systems

    SciTech Connect (OSTI)

    Sacks, I.J.; Ashmore, B.C.; Champney, J.M.; Alesso, H.P.

    1983-06-01

    This report provides preliminary results generated by a Digraph Matrix Analysis (DMA) for a Systems Interaction analysis performed on the Safety Injection System of the Tennessee Valley Authority Watts Bar Nuclear Power Plant. An overview of DMA is provided along with a brief description of the computer codes used in DMA.

  12. Safety review advisor

    SciTech Connect (OSTI)

    Boshers, J.A.; Alguindigue, I.E.; Uhrig, R.E. (Tennessee Univ., Knoxville, TN (USA). Dept. of Nuclear Engineering); Burnett, C.G. (Tennessee Valley Authority, Knoxville, TN (USA))

    1989-01-01

    The University of Tennessee's Nuclear Engineering Department, in cooperation with the Tennessee Valley Authority (TVA), is evaluating the feasibility of utilizing an expert system to aid in 10CFR50.59 evaluations. This paper discusses the history of 10CFR50.59 reviews, and details the development approach used in the construction of a prototype Safety Review Advisor (SRA). The goals for this expert system prototype are to (1) aid the engineer in the evaluation process by directing his attention to the appropriate critical issues, (2) increase the efficiency, consistency, and thoroughness of the evaluation process, and (3) provide a foundation of appropriate Safety Analysis Report (SAR) references for the reviewer. 6 refs., 2 figs.

  13. HEALTH EFFECTS OF THE NUCLEAR ACCIDENT AT THREE MILE ISLAND

    E-Print Network [OSTI]

    Fabrikant, J.I.

    2010-01-01

    In) Symposium on Nuclear Reactor Safety: Perspective. Awe learned about nuclear reactor safety and health from the

  14. Activities in Support of Continuing the Service of Nuclear Power Plant Safety-Related Concrete Structures

    SciTech Connect (OSTI)

    Naus, Dan J

    2014-01-01

    Nuclear power plant (NPP) concrete structures are described. In-service inspection and testing requirements in the U.S. are summarized. The license renewal process in the U.S. is outlined and its current status provided. Operating experience related to performance of the concrete structures is presented. Basic components of a program to manage aging of the concrete structures are identified and described: (1) Degradation mechanisms, damage models, and material performance; (2) Assessment and remediation: i.e., component selection, in- service inspection, non-destructive examinations, and remedial actions; and (3) Estimation of performance at present or some future point in time: i.e., application of structural reliability theory to the design and optimization of in-service inspection/maintenance strategies, and determination of the effects of degradation on plant risk. Finally, areas are noted where additional research would be of benefit to aging management of nuclear power plant concrete structures.

  15. Facility Safety | Department of Energy

    Office of Environmental Management (EM)

    improved DOE program and field implementation of nuclear safety management programs and fire protection and natural phenomena hazard control requirements. Nuclear facility program...

  16. e-Docs # 4179757 Security of Nuclear Substances

    E-Print Network [OSTI]

    Chan, Hue Sun

    Nuclear Security Division Canadian Nuclear Safety Commission Toronto Outreach October 30, 2013 Raphael Duguay, Security Advisor Nuclear Security Division Canadian Nuclear Safety Commission #12;2Canadian Nuclear Safety Commission / Commission canadienne de sûreté nucléaire Presentation outline

  17. Preliminary Safety Design RM

    Office of Environmental Management (EM)

    Preliminary Safety Design Review Module March 2010 CD-0 O 0 OFFICE OF Pr C CD-1 F ENVIRO Standard R reliminar Rev Critical Decis CD-2 M ONMENTAL Review Plan ry Safety view Module...

  18. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    SciTech Connect (OSTI)

    Carvo, Alan E.

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  19. 10 CFR Part 830 Nuclear Safety Technical Positions | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative FuelsofProgram:Y-12 Beta-3 Racetracks Y-12 Beta-3of/ The Office of10 CFR 851Nuclear10

  20. Long-Term Planning for Nuclear Energy Systems Under Deep Uncertainty

    E-Print Network [OSTI]

    Kim, Lance Kyungwoo

    2011-01-01

    iii 3 Nuclear Fuel Cycle Safety 3.1 Overview . . . . .system performance related to nuclear safety. Factors thatof safety risk posed by various nuclear facilities . . . . .

  1. DEVELOPMENT OF ELECTRO-OPTICAL INSTRUMENTATION FOR REACTOR SAFETY STUDIES

    E-Print Network [OSTI]

    Turko, B.T.

    2010-01-01

    and accuracy needed for nuclear safety assessment. TheThe assessment of the safety of nuclear reactori two-phaseNuclear Science DEVELOPMENT OF ELECTRO-OPTICAL INSTRUMENTATION FOR REACTOR SAFETY

  2. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    SciTech Connect (OSTI)

    Weiss, A. J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately.

  3. A case study on the influence of THM coupling on the near field safety of a spent fuel repository in sparsely fractured granite

    E-Print Network [OSTI]

    Nguyen, T.S.

    2009-01-01

    Jonny Rutqvist 7 Canadian Nuclear Safety Commission, Ottawa,completion: the Canadian Nuclear Safety Commission (CNSC);repository is called a safety case (Nuclear Energy Agency,

  4. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    the possible risk from nuclear power . it . is sufficient tothe Cancer Risk Due to Nuclear-Electric Power Generation",of Accident Risks in U.S. Commercial Nuclear Power Plants",

  5. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    in U. S. Conunercial Nuclear Power Plants", Report WASH-Related Standards for Nuclear Power Plants," by A.V. NeroResponse Planning for Nuclear Power Plants in California,"

  6. Analysis of Flow in Pilot Operated Safety and Relief Valve of Nuclear Reactor Coolant System

    SciTech Connect (OSTI)

    Kwon, Soon-Bum; Lee, Dong-Won; Kim, In-Goo; Ahn, Hyung-Joon; Kim, Hho-Jung

    2004-07-01

    When the POSRV equipped in a nuclear power plant opens in instant by a failure in coolant system of PWR, a moving shock wave generates, and propagates downstream of the valve, inducing a complicated unsteadiness. The moving shock wave may exert severe load to the structure. In this connection, a method of gradual opening of the valve is used to reduce the load acting on the wall at the downstream of the POSRV. In the present study, experiments and calculations are performed to investigate the detail unsteady flow at the various pipe units and the effect of valve opening time on the flow downstream of the valve. In calculation by using of air as working fluid, 2-dimensional, unsteady compressible Navier-Stokes equations are solved by finite volume method. It was found that when the incident shock wave passes through the pipe unit, it may experience diffraction, reflection and interaction with a vortex. Furthermore, the geometry of the pipe unit affects the reflection type of shock wave and changes the load acting on the wall of pipe unit. It was also turned out that the maximum force acting on the wall of the pipe unit becomes in order of T-junction, 108 deg. elbow and branch in magnitude, respectively. And, the results obtained that show that the rapid pressure rise due to the moving shock wave by instant POSRV valve opening is attenuated by employing the gradual opening. (authors)

  7. Criticality Safety Study of UF6and UO2F2in 8-in. Inner Diameter Piping

    SciTech Connect (OSTI)

    Elam, K.R.

    2003-10-07

    The purpose of this report is to provide an evaluation of the criticality safety aspects of using up to 8-in.-inner-diameter (ID) piping as part of a system to monitor the {sup 235}U enrichment in uranium hexafluoride (UF{sub 6}) gas both before and after an enrichment down-blending operation. The evaluated operation does not include the blending stage but includes only the monitors and the piping directly associated with the monitors, which are in a separate room from the blending operation. There are active controls in place to limit the enrichment of the blended UF{sub 6} to a maximum of 5 weight percent (wt%) {sup 235}U. Under normal operating conditions of temperature and pressure, the UF{sub 6} will stay in the gas phase and criticality will not be credible. The two accidents of concern are solidification of the UF{sub 6} along with some hydrofluoric acid (HF) and water or moisture ingress, which would cause the UF{sub 6} gas to react and form a hydrated uranyl fluoride (UO{sub 2}F{sub 2}) solid or solution. Of these two types of accidents, the addition of water and formation of UO{sub 2}F{sub 2} is the most reactive scenario and thus limits related to UO{sub 2}F{sub 2} will bound the limits related to UF{sub 6}. Two types of systems are included in the monitoring process. The first measures the enrichment of the approximately 90 wt% enriched UF{sub 6} before it is blended. This system uses a maximum 4-in.-(10.16-cm-) ID pipe, which is smaller than the 13.7-cm-cylinder-diameter subcritical limit for UO{sub 2}F{sub 2} solution of any enrichment as given in Table 1 of American National Standard ANSI/ANS-8.1.1 Therefore, this system poses no criticality concerns for either accident scenario. The second type of system includes two enrichment monitors for lower-enriched UF{sub 6}. One monitors the approximately 1.5 wt% enriched UF{sub 6} entering the blending process, and the second monitors the approximately 5 wt% enriched UF{sub 6} coming out of the blending process. Both use a maximum 8-in.-(20.32-cm-) ID piping, where the length of the larger ID piping is approximately 9.5 m. This diameter of piping is below the 26.6-cm-cylinder-diameter subcritical limit for 5 wt% enriched UO{sub 2}F{sub 2} solutions as given in Table 6 of ANSI/ANS-8.1. Therefore, for up to 5 wt% enriched UF{sub 6}, this piping does not present a criticality concern for either accident scenario. Calculations were performed to determine the enrichment level at which criticality could become a concern in these 8-in.-ID piping sections. Both unreflected and fully water-reflected conditions were considered.

  8. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

    SciTech Connect (OSTI)

    Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

    2011-02-01

    This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

  9. Criticality Model

    SciTech Connect (OSTI)

    A. Alsaed

    2004-09-14

    The ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003) presents the methodology for evaluating potential criticality situations in the monitored geologic repository. As stated in the referenced Topical Report, the detailed methodology for performing the disposal criticality analyses will be documented in model reports. Many of the models developed in support of the Topical Report differ from the definition of models as given in the Office of Civilian Radioactive Waste Management procedure AP-SIII.10Q, ''Models'', in that they are procedural, rather than mathematical. These model reports document the detailed methodology necessary to implement the approach presented in the Disposal Criticality Analysis Methodology Topical Report and provide calculations utilizing the methodology. Thus, the governing procedure for this type of report is AP-3.12Q, ''Design Calculations and Analyses''. The ''Criticality Model'' is of this latter type, providing a process evaluating the criticality potential of in-package and external configurations. The purpose of this analysis is to layout the process for calculating the criticality potential for various in-package and external configurations and to calculate lower-bound tolerance limit (LBTL) values and determine range of applicability (ROA) parameters. The LBTL calculations and the ROA determinations are performed using selected benchmark experiments that are applicable to various waste forms and various in-package and external configurations. The waste forms considered in this calculation are pressurized water reactor (PWR), boiling water reactor (BWR), Fast Flux Test Facility (FFTF), Training Research Isotope General Atomic (TRIGA), Enrico Fermi, Shippingport pressurized water reactor, Shippingport light water breeder reactor (LWBR), N-Reactor, Melt and Dilute, and Fort Saint Vrain Reactor spent nuclear fuel (SNF). The scope of this analysis is to document the criticality computational method. The criticality computational method will be used for evaluating the criticality potential of configurations of fissionable materials (in-package and external to the waste package) within the repository at Yucca Mountain, Nevada for all waste packages/waste forms. The criticality computational method is also applicable to preclosure configurations. The criticality computational method is a component of the methodology presented in ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003). How the criticality computational method fits in the overall disposal criticality analysis methodology is illustrated in Figure 1 (YMP 2003, Figure 3). This calculation will not provide direct input to the total system performance assessment for license application. It is to be used as necessary to determine the criticality potential of configuration classes as determined by the configuration probability analysis of the configuration generator model (BSC 2003a).

  10. Nuclear Systems Technology | Nuclear Science | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Cycle Systems Criticality Safety Irradiation Experiment Development and Execution Robotics & Remote Systems Engineering and Applications Thermal & Hydraulic Experiments &...

  11. Senior Technical Safety Manager Qualification Program Self-Assessment...

    Office of Environmental Management (EM)

    Senior Technical Safety Manager Qualification Program Self-Assessment - Chief of Nuclear Safety Senior Technical Safety Manager Qualification Program Self-Assessment - Chief of...

  12. Advanced Modeling and Evaluation of the Response of Base-Isolated Nuclear Facility Structures to Vertical Earthquake Excitation

    E-Print Network [OSTI]

    Keldrauk, Eric Scott

    2012-01-01

    until structural and nuclear safety inspections could beSeismic analysis of safety-related nuclear structures anddamage to nuclear equipment or safety related components,

  13. Nuclear Energy Governance and the Politics of Social Justice: Technology, Public Goods, and Redistribution in Russia and France

    E-Print Network [OSTI]

    Grigoriadis, Theocharis N

    2009-01-01

    science and 4. Nuclear and radiation safety. 10 Thisof the nuclear fuel cycle as well as safety measures drivewhy safety has been the central priority of French nuclear

  14. Criticality Safety | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuels DataEnergy Webinar:IAbout Us » ContactCounty Aims to Save withManagers

  15. As officials in Japan deal with the accumulation of radioactive seawater near the devastated Fukushima Daiichi nuclear power plant in the wake of last month's

    E-Print Network [OSTI]

    Danon, Yaron

    Fukushima Daiichi nuclear power plant in the wake of last month's earthquake and tsunami, the U.S. Department of Energy is investing in fundamental research it hopes can be used to build safer nuclear reactors and avoid reactor emergencies. The department's Nuclear Criticality Safety Program (NCSP

  16. Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

    SciTech Connect (OSTI)

    Krass, A.W.

    2005-12-19

    This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

  17. Interdisciplinary: Industrial Hygienist/Safety Engineer

    Broader source: Energy.gov [DOE]

    This position is located in the Office of Worker Safety and Health Assessments that conducts assessments to provide critical feedback and objective information on occupational safety and health...

  18. Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).

    SciTech Connect (OSTI)

    Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

    2010-09-01

    This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

  19. RADIOLOGICAL HEALTH AND RELATED STANDARDS FOR NUCLEAR POWER PLANTS. VOLUME 2 OF HEALTH AND SAFETY IMPACTS OF NUCLEAR, GEOTHERMAL, AND FOSSIL-FUEL ELECTRIC GENERATION IN CALIFORNIA

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01

    Report LBL-5287. "Power Plant Reliability-Availability andConunercial Nuclear Power Plants", Report WASH-1400 (NUREG-Standards for Nuclear Power Plants," by A.V. Nero and Y.C.

  20. The Decline and Death of Nuclear Power

    E-Print Network [OSTI]

    Melville, Jonathan

    2013-01-01

    1. The International Nuclear Safety Advisory Group (Safetystrict safety regulations regarding containment of nuclearnuclear power plant poses is greater than any other source of energy, and no safety

  1. Safety Advisor (Nuclear Engineer)

    Broader source: Energy.gov [DOE]

    WHO MAY APPLY: Current or former (competitive and excepted service) Federal civilian employees; Veterans discharged under honorable conditions (this means an honorable or general discharge), AND...

  2. Safety Advisor (Nuclear Engineer)

    Broader source: Energy.gov [DOE]

    **This is an Excepted Service position. This appointment will not confer Competitive Service career-conditional or career tenure status. This means that if you are selected, you would have to...

  3. Nuclear Safety Regulatory Framework

    Energy Savers [EERE]

    of workers, the public and the environment. To achieve this, the department ensures that radiation exposures to its workers and the public and releases to the environment are...

  4. Nuclear Safety Analysis Reports

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1992-04-30

    Cancels DOE O 5481.1B; paragraphs 7b(3), 7e(3) & 8c of DOE O 5480.6; and 51, 7b(3), 7b(4), 7e(3), 8a & 8h of DOE O 5480.5.

  5. Nuclear Safety Regulatory Framework

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested Parties -DepartmentAvailable forSite |n t e OfficeResearchDepartment of Energy

  6. Radioisotope Power System Delivery, Ground Support and Nuclear Safety Implementation: Use of the Multi-Mission Radioisotope Thermoelectric Generator for the NASA's Mars Science Laboratory

    SciTech Connect (OSTI)

    S.G. Johnson; K.L. Lively; C.C. Dwight

    2014-07-01

    Radioisotope power systems have been used for over 50 years to enable missions in remote or hostile environments. They are a convenient means of supplying a few milliwatts up to a few hundred watts of useable, long-term electrical power. With regard to use of a radioisotope power system, the transportation, ground support and implementation of nuclear safety protocols in the field is a complex process that requires clear identification of needed technical and regulatory requirements. The appropriate care must be taken to provide high quality treatment of the item to be moved so it arrives in a condition to fulfill its missions in space. Similarly it must be transported and managed in a manner compliant with requirements for shipment and handling of special nuclear material. This presentation describes transportation, ground support operations and implementation of nuclear safety and security protocols for a radioisotope power system using recent experience involving the Multi-Mission Radioisotope Thermoelectric Generator for National Aeronautics and Space Administration’s Mars Science Laboratory, which launched in November of 2011.

  7. High-altitude-nuclear electromagnetic pulse (HEMP) environment simulation public health and safety considerations. Technical report, 1 Oct 89-31 May 91

    SciTech Connect (OSTI)

    Casey, K.

    1992-03-01

    The existence of electromagnetic fields external to the working volumes of high-altitude nuclear electromagnetic pulse (HEMP) environment simulators has raised both environmental and public-health concerns regarding the safety of HEMP environment simulator operations. This report contains a review of what HEMP is, what its effects on defense systems are, and why and how HEMP environment simulation testing is conducted. The state of present knowledge concerning the external simulator fields and their possible effects on biological and electronic systems is summarized. Research initiatives are identified to aid in answering the most important questions regarding the continued environmental safety of HEMP simulator operations. These initiatives are intended to support (1) development of options for modification and/or relocation of HEMP environment simulator facilities and (2) determination of safe exposure levels for biological and electronic systems. Recommendations for specific DoD actions are given.

  8. ICSBEP Benchmarks For Nuclear Data Applications

    SciTech Connect (OSTI)

    Briggs, J. Blair [Idaho National Engineering and Environmental Laboratory (United States)

    2005-05-24

    The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the United States Department of Energy. The ICSBEP became an official activity of the Organization for Economic Cooperation and Development (OECD) -- Nuclear Energy Agency (NEA) in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Serbia and Montenegro (formerly Yugoslavia), Kazakhstan, Spain, Israel, Brazil, Poland, and the Czech Republic are now participating. South Africa, India, China, and Germany are considering participation. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled ''International Handbook of Evaluated Criticality Safety Benchmark Experiments.'' The 2004 Edition of the Handbook contains benchmark specifications for 3331 critical or subcritical configurations that are intended for use in validation efforts and for testing basic nuclear data. New to the 2004 Edition of the Handbook is a draft criticality alarm / shielding type benchmark that should be finalized in 2005 along with two other similar benchmarks. The Handbook is being used extensively for nuclear data testing and is expected to be a valuable resource for code and data validation and improvement efforts for decades to come. Specific benchmarks that are useful for testing structural materials such as iron, chromium, nickel, and manganese; beryllium; lead; thorium; and 238U are highlighted.

  9. TWRS safety program plan

    SciTech Connect (OSTI)

    Calderon, L.M., Westinghouse Hanford

    1996-08-01

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

  10. New Nuclear Reaction Flow during r-Process Nucleosynthesis in Supernovae: Critical Role of Light Neutron-Rich Nuclei

    E-Print Network [OSTI]

    M. Terasawa; K. Sumiyoshi; T. Kajino; G. J. Mathews; I. Tanihata

    2001-07-19

    We study the role of light neutron-rich nuclei during r-process nucleosynthesis in supernovae. Most previous studies of the r-process have concentrated on the reaction flow of heavy unstable nuclei. Although the nuclear reaction network includes a few thousand heavy nuclei, only limited reaction flow through light-mass nuclei near the stability line has been used in those studies. However, in a viable scenario of the r-process in neutrino-driven winds, the initial condition is a high-entropy hot plasma consisting of neutrons, protons, and electron-positron pairs experiencing an intense flux of neutrinos. In such environments light-mass nuclei as well as heavy nuclei are expected to play important roles in the production of seed nuclei and r-process elements. Thus, we have extended our fully implicit nuclear reaction network so that it includes all nuclei up to the neutron drip line for Z $ \\leq 10$, in addition to a larger network for Z $ \\geq 10$. In the present nucleosynthesis study, we utilize a wind model of massive SNeII explosions to study the effects of this extended network. We find that a new nuclear-reaction flow path opens in the very light neutron-rich region. This new nuclear reaction flow can change the final heavy-element abundances by as much as an order of magnitude.

  11. RISMC ADVANCED SAFETY ANALYSIS WORKING PLAN – FY 2015 – FY 2019

    SciTech Connect (OSTI)

    Szilard, Ronaldo H; Smith, Curtis L

    2014-09-01

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

  12. A Joint Interview with Professor Joonhong Ahn and Professor Cathryn Carson on Nuclear Waste Management: a Technical and Social Problem

    E-Print Network [OSTI]

    Chowdhary, Harshika; Gill, Manraj; Kim, Juwon; McGuinness, Philippa; Miller, Daniel; Nuckolls, Kevin

    2015-01-01

    Traditionally, the nuclear safety has been established basedNuclear engineers have to consider the social aspects and the safety

  13. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2013-06-21

    DOE-STD-1104 contains the Department's method and criteria for reviewing and approving nuclear facility's documented safety analysis (DSA). This review and approval formally document the basis for DOE, concluding that a facility can be operated safely in a manner that adequately protects workers, the public, and the environment. Therefore, it is appropriate to formally require implementation of the review methodology and criteria contained in DOE-STD-1104.

  14. Spin-On for the Renaissance? The Current State of China's Nuclear Industry

    E-Print Network [OSTI]

    Yuan, Jing-dong

    2010-01-01

    cal innovation, nuclear safety, and strategic plan- ning tonuclear technology trans- fers and facility and materials safety.safety systems, is “the safest, most advanced, yet proven nuclear

  15. SaCS: A Method and a Pattern Language for the Development of Conceptual Safety Designs

    E-Print Network [OSTI]

    Stølen, Ketil

    , and nuclear reactor protection systems are examples of safety critical systems from different industrial.D.) in Computer Science June 2014 #12;ii #12;Abstract Flight control systems, railway interlocking systems. In this thesis we present a method and a pattern language called Safe Control Systems (SaCS) for development

  16. Audit Report The Procurement of Safety Class/Safety-Significant Items at the Savannah River Site

    SciTech Connect (OSTI)

    None

    2009-04-01

    The Department of Energy operates several nuclear facilities at its Savannah River Site, and several additional facilities are under construction. This includes the National Nuclear Security Administration's Tritium Extraction Facility (TEF) which is designated to help maintain the reliability of the U.S. nuclear stockpile. The Mixed Oxide Fuel Fabrication Facility (MOX Facility) is being constructed to manufacture commercial nuclear reactor fuel assemblies from weapon-grade plutonium oxide and depleted uranium. The Interim Salt Processing (ISP) project, managed by the Office of Environmental Management, will treat radioactive waste. The Department has committed to procuring products and services for nuclear-related activities that meet or exceed recognized quality assurance standards. Such standards help to ensure the safety and performance of these facilities. To that end, it issued Departmental Order 414.1C, Quality Assurance (QA Order). The QA Order requires the application of Quality Assurance Requirements for Nuclear Facility Applications (NQA-1) for nuclear-related activities. The NQA-1 standard provides requirements and guidelines for the establishment and execution of quality assurance programs during the siting, design, construction, operation, and decommissioning of nuclear facilities. These requirements, promulgated by the American Society of Mechanical Engineers, must be applied to 'safety-class' and 'safety-significant' structures, systems and components (SSCs). Safety-class SSCs are defined as those necessary to prevent exposure off site and to protect the public. Safety-significant SSCs are those whose failure could irreversibly impact worker safety such as a fatality, serious injury, or significant radiological or chemical exposure. Due to the importance of protecting the public, workers, and environment, we initiated an audit to determine whether the Department of Energy procured safety-class and safety-significant SSCs that met NQA-1 standards at the Savannah River Site. Our review disclosed that the Department had procured and installed safety-class and safety-significant SSCs that did not meet NQA-1 quality standards. Specifically, we identified multiple instances in which critical components did not meet required quality and safety standards. For example: (1) Three structural components were procured and installed by the prime contractor at Savannah River during construction of the MOX Facility that did not meet the technical specifications for items relied on for safety. These substandard items necessitated costly and time consuming remedial action to, among other things, ensure that nonconforming materials and equipment would function within safety margins; (2) In six instances, items used in the construction of TEF failed to satisfy quality standards. In one of these situations, operating procedures had to be modified to ensure that the problem item did not compromise safety; and (3) Finally, at the ISP, one component that did not meet quality standards was procured. The failure of the item could have resulted in a spill of up to 15,000 gallons of high-level radioactive waste. Based on an extensive examination of relevant internal controls and procurement practices, we concluded that these failures were attributable to inadequate attention to quality assurance at Savannah River. Simply put, Departmental controls were not adequate to prevent and/or detect quality assurance problems. For example, Federal and prime contractor officials did not expressly require that subcontractors or lower-tiered vendors comply with quality assurance requirements. Additionally, management did not effectively communicate quality assurance concerns between the several Departmental program elements operating at Savannah River. The procurement and installation of these nonconforming components resulted in cost increases. For example, as of October 2008, the MOX Facility had incurred costs of more than $680,000 due to problems associated with the procurement of $11 million of nonconforming safety-class reinforcing steel.

  17. Criticality calculations with MCNP{sup TM}: A primer

    SciTech Connect (OSTI)

    Mendius, P.W. [ed.; Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

    1994-08-01

    The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.

  18. Gordon wins NNSA Safety Professional

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Year award March 12, 2009 LOS ALAMOS, New Mexico, March 12, 2009-Laboratory Chief Electrical Safety Officer Lloyd Gordon received the 2008 National Nuclear Security...

  19. Financing arrangements and industrial organisation for new nuclear build in electricity markets

    E-Print Network [OSTI]

    Finon, Dominique; Roques, Fabien A.

    acceptability on plant sitting and nuclear waste management. 3 From this perspective, governments and regulatory and safety agencies have a critical role to play in the setting of clear and consistent procedures for licensing design and authorisation... cycles could undermine the commitment to nuclear power and the development of nuclear waste disposal facilities; and iii) the construction and operation risks associated with the necessary “re-learning” of the technology. Moreover, the large size of a...

  20. Review guidelines for software languages for use in nuclear power plant safety systems: Final report. Revision 1

    SciTech Connect (OSTI)

    Hecht, M.; Decker, D.; Graff, S.; Green, W.; Lin, D.; Dinsmore, G.; Koch, S.

    1997-10-01

    Guidelines for the programming and auditing of software written in high level languages for safety systems are presented. The guidelines are derived from a framework of issues significant to software safety which was gathered from relevant standards and research literature. Language-specific adaptations of these guidelines are provided for the following high level languages: Ada83 and Ada95; C and C++; International Electrochemical Commission (IEC) Standard 1131-3 Ladder Logic, Sequential Function Charts, Structured Text, and Function Block Diagrams; Pascal; and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.