National Library of Energy BETA

Sample records for nuclear criticality safety

  1. Autoclave nuclear criticality safety analysis

    SciTech Connect (OSTI)

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

    1991-12-31

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

  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. 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.

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

    SciTech Connect (OSTI)

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

    2009-11-01

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

  6. 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. Proceedings of the Nuclear Criticality Technology Safety Workshop

    SciTech Connect (OSTI)

    Rene G. Sanchez

    1998-04-01

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

  8. Tank waste remediation system nuclear criticality safety program management review

    SciTech Connect (OSTI)

    BRADY RAAP, M.C.

    1999-06-24

    This document provides the results of an internal management review of the Tank Waste Remediation System (TWRS) criticality safety program, performed in advance of the DOE/RL assessment for closure of the TWRS Nuclear Criticality Safety Issue, March 1994. Resolution of the safety issue was identified as Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-40-12, due September 1999.

  9. Criticality Safety | Department of Energy

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

    Services » Nuclear Safety » Criticality Safety Criticality Safety Nuclear Safety Basis The Nuclear Facility Safety Program establishes and maintains the DOE requirements for nuclear criticality safety. The DOE detailed requirements for criticality safety are contained in Section 4.3 of the DOE Order 420.1,Facility Safety. Criticality safety requirements are based on the documented safety analysis required by 10 CFR 830, Subpart B. Related Links 10 CFR 830, Nuclear Safety Management American

  10. 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.

  11. 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.

  12. Y-12's 1958 nuclear criticality accident and increased safety...

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

    accident and increased safety - 1958 brought accidents, more safety The first X-ray machine was brought to Y-12 in February, 1949. It was a 1,000 KV system installed in Building...

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

    SciTech Connect (OSTI)

    Huffer, J.E.

    1997-04-01

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

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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.

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

    SciTech Connect (OSTI)

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

    2011-09-01

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

  20. Deep geologic burial of spent nuclear fuel: Is criticality a public health and safety issue?

    SciTech Connect (OSTI)

    McLaughlin, T.P.

    1996-12-31

    While the answer to the question posed in the title to this paper may never be complete, there is evidence that suggests that the technical answer is {open_quotes}no.{close_quotes} Certainly there will likely be vigorous public policy discussions as to the acceptability of criticality events at indeterminate times in the future even if the technical arguments for acceptably low risk are compelling. This paper attempts to further the technical discussions of criticality events associated with geologic disposal of fissile material being considered acceptably low risks to future inhabitants. Current U.S. regulations governing the deep geologic disposal of materials that may be capable of achieving the critical state are found in 10 CFR 60 of the Code of Federal Regulations. The pertinent paragraph, 60.131(b)(7), states: {open_quotes}Criticality control. All systems for processing, transporting, handling, storage, retrieval, emplacement, and isolation of radioactive waste shall be designed to ensure that a nuclear criticality accident is not possible unless at least two unlikely, independent, and concurrent or sequential changes have occurred in the conditions essential to nuclear criticality safety. Each system shall be designed for criticality safety under normal and accident conditions. The calculated effective multiplication factor (k{sub eff}) must be sufficiently below unity to show at least a 5% margin, after allowance for the bias in the method of calculation and the uncertainty in the experiments used to validate the method of calculation.{close_quotes}

  1. 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.

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

    Energy Savers [EERE]

    5-99 September 1999 DOE STANDARD GUIDANCE FOR NUCLEAR CRITICALITY SAFETY ENGINEER TRAINING AND QUALIFICATION U.S. Department of Energy AREA SAFT Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS This document has been reproduced from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U.S. Department of Energy, (800) 473-4375, fax: (301) 903-9823. Available to the public from the

  3. 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.

  4. Tank waste remediation system nuclear criticality safety inspection and assessment plan

    SciTech Connect (OSTI)

    VAIL, T.S.

    1999-04-06

    This plan provides a management approved procedure for inspections and assessments of sufficient depth to validate that the Tank Waste Remediation System (TWRS) facility complies with the requirements of the Project Hanford criticality safety program, NHF-PRO-334, ''Criticality Safety General, Requirements''.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

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

    SciTech Connect (OSTI)

    Lee, B.L. Jr.; D`Aquila, D.M.

    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.

  10. CRITICALITY SAFETY TRAINING AT FLUOR HANFORD (FH)

    SciTech Connect (OSTI)

    TOFFER, H.

    2005-05-02

    The Fluor Hanford Criticality Safety engineers are extensively trained. The objectives and requirements for training are derived from Department of Energy (DOE) and American National Standards Institute/American Nuclear Society Standards (ANSI/ANS), and are captured in the Hanford Criticality Safety Program manual, HNF-7098. Qualification cards have been established for the general Criticality Safety Engineer (CSE) analyst, CSEs who support specific facilities, and for the facility Criticality Safety Representatives (CSRs). Refresher training and continuous education in the discipline are emphasized. Weekly Brown Bag Sessions keep the criticality safety engineers informed of the latest developments and historic perspectives.

  11. Nuclear Safety | Department of Energy

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

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

  12. Nuclear Safety Enforcement Documents | Department of Energy

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

    Nuclear Safety Enforcement Documents Nuclear Safety Enforcement Documents Documents Available for Download April 28, 2016 Savannah River Nuclear Solutions, LLC, Consent Order NCO-2016-01 Nuclear Safety Enforcement Consent Order issued to Savannah River Nuclear Solutions, LLC relating to nuclear criticality safety infractions that occurred at the Savannah River Site February 19, 2016 Preliminary Notice of Violation, Los Alamos National Security, LLC Nuclear Safety Enforcement Preliminary Notice

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. Nuclear Safety Regulatory Framework

    Energy Savers [EERE]

    Department of Energy Nuclear Safety Regulatory Framework DOE's Nuclear Safety Enabling Legislation Regulatory Enforcement & Oversight Regulatory Governance Atomic Energy Act 1946 Atomic Energy Act 1954 Energy Reorganization Act 1974 DOE Act 1977 Authority and responsibility to regulate nuclear safety at DOE facilities 10 CFR 830 10 CFR 835 10 CFR 820 Regulatory Implementation Nuclear Safety Radiological Safety Procedural Rules ISMS-QA; Operating Experience; Metrics and Analysis Cross Cutting

  18. AGING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2004-09-10

    The purpose of this design calculation is to revise and update the previous criticality calculation for the Aging Facility (documented in BSC 2004a). This design calculation will also demonstrate and ensure that the storage and aging operations to be performed in the Aging Facility meet the criticality safety design criteria in the ''Project Design Criteria Document'' (Doraswamy 2004, Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''SNF Aging System Description Document'' (BSC [Bechtel SAIC Company] 2004f, p. 3-12). The scope of this design calculation covers the systems and processes for aging commercial spent nuclear fuel (SNF) and staging Department of Energy (DOE) SNF/High-Level Waste (HLW) prior to its placement in the final waste package (WP) (BSC 2004f, p. 1-1). Aging commercial SNF is a thermal management strategy, while staging DOE SNF/HLW will make loading of WPs more efficient (note that aging DOE SNF/HLW is not needed since these wastes are not expected to exceed the thermal limits form emplacement) (BSC 2004f, p. 1-2). The description of the changes in this revised document is as follows: (1) Include DOE SNF/HLW in addition to commercial SNF per the current ''SNF Aging System Description Document'' (BSC 2004f). (2) Update the evaluation of Category 1 and 2 event sequences for the Aging Facility as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004c, Section 7). (3) Further evaluate the design and criticality controls required for a storage/aging cask, referred to as MGR Site-specific Cask (MSC), to accommodate commercial fuel outside the content specification in the Certificate of Compliance for the existing NRC-certified storage casks. In addition, evaluate the design required for the MSC that will accommodate DOE SNF/HLW. This design calculation will achieve the objective of providing the criticality safety results to support the preliminary design of the Aging Facility. As the ongoing design evolution remains fluid, the results from this design calculation should be evaluated for applicability to any new or modified design. Consequently, the results presented in this document are limited to the current design. The information contained in this document was developed by Environmental and Nuclear Engineering and is intended for the use of Design and Engineering in its work regarding the various criticality related activities performed in the Aging Facility. Yucca Mountain Project personnel from Environmental and Nuclear Engineering should be consulted before the use of the information for purposes other than those stated herein or use by individuals other than authorized personnel in Design and Engineering.

  19. 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.

  20. 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).

  1. CRAD, NNSA- Criticality Safety (CS)

    Office of Energy Efficiency and Renewable Energy (EERE)

    CRAD for Criticality Safety (CS). 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.

  2. 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.

  3. 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.

  4. Criticality Safety Training

    Energy Science and Technology Software Center (OSTI)

    2002-12-01

    CST is a web-based training program designed to help the user to safely access and work in areas where fissionable nuclear materials may be present.

  5. Guidelines for Preparing Criticality Safety Evaluations at Department...

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

    7-2007, Guidelines for Preparing Criticality Safety Evaluations at Department of Energy Non-Reactor Nuclear Facilities by Diane Johnson This standard provides a framework for...

  6. DOE-STD-1173-2003; Criticality Safety Functional Area Qualification...

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

    ... 15. Criticality safety personnel shall demonstrate a working level knowledge of the Nuclear Safety Management Rule, 10CFR830, requirements related to Unreviewed Safety Questions ...

  7. Nuclear Safety | Department of Energy

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

    Nuclear Safety Nuclear Safety The Nuclear Safety Program mission is to support the design, construction, operation, and deactivation and decommissioning of the Paducah and Portsmouth nuclear facilities in a manner that ensures adequate protection of workers, the public, and the environment. Major Responsibilities: Establish and implement nuclear safety requirements that utilize national consensus (or other government) standards or applicable external agency regulations (Nuclear Regulatory

  8. 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.

  9. 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.

  10. 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

  11. Office of Nuclear Safety | Department of Energy

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

    Nuclear Safety Office of Nuclear Safety Mission The Office of Nuclear Safety establishes nuclear safety requirements and expectations for the Department to ensure protection of ...

  12. Defense Nuclear Facility Safety Board

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

    8, 2014 Defense Nuclear Facility Safety Board Defense Nuclear Facility Safety Board (DNSFB) Vice Chairwoman Jesse Roberson visited and toured the WIPP site this week. While...

  13. 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.

  14. Nuclear Energy Safety Technologies

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

    Safety Technologies - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced

  15. WIPP Documents - Nuclear Safety

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

    Nuclear Safety DOE/WIPP-07-3372, Revision 4 WIPP Documented Safety Analysis Approved November 2013 The Documented Safety Analysis addresses all hazards (both radiological and nonradiological) and the controls necessary to provide adequate protection to the public, workers, and the environment. The WIPP DSA demonstrates the extent to which the Waste Isolation Pilot Plant can be operated safely with respect to workers, the public, and the environment. DOE/WIPP-07-3373, Revision 4 WIPP Technical

  16. 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.

  17. 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

  18. 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.

  19. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility. However, it is anticipated that design changes to the facility layout will have little or no impact on the criticality results and/or conclusions presented in this document. This calculation is subject to the ''Quality Assurance Requirements and Description'' (DOE 2004 [DIRS 171539]) because the CHF is included in the Q-List (BSC 2005 [DIRS 171190], p. A-3) as an item important to safety. This calculation is prepared in accordance with AP-3.12Q, ''Design Calculations and Analyses'' [DIRS 168413].

  20. 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.; Pryor, W.A.

    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.

  1. Safety Culture in Nuclear Installations

    Broader source: Energy.gov [DOE]

    IAEA-TECDOC-1329 Safety Culture in Nuclear Installations, Guidance for use in the Enhancement of Safety Culture, International Atomic Energy Agency IAEA, December 2002.

  2. 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.

  3. 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.

  4. Pilot Project: Nuclear Safety Information Dashboard | Department...

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

    Pilot Project: Nuclear Safety Information Dashboard Pilot Project: Nuclear Safety Information Dashboard The Nuclear Safety Information (NSI) Dashboard provides a new user interface ...

  5. 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.

  6. 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)

  7. 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.

  8. Progress and goals for INMM ASC N15 consensus standard ""Administrative practices for the determination and reporting of results of non-destructive assay measurements of nuclear material in situ for safeguards nuclear criticality safety and other purposes

    SciTech Connect (OSTI)

    Bracken, David S; Lamb, Frank W

    2009-01-01

    This paper will discuss the goals and progress to date on the development of INMM Accredited Standard Committee (ASC) N15 consensus standard Administrative Practices for the Determination and Reporting of Results of Non-Destructive Assay Measurements of Nuclear Material in situ for Safeguards, Nuclear Criticality Safety, and Other Purposes. This standard will define administrative practices in the areas of data generation and reporting of NDA assay of holdup deposits with consideration of the stakeholders of the reported results. These stakeholders may include nuclear material accounting and safeguards, nuclear criticality safety, waste management, health physics, facility characterization, authorization basis, radiation safety, and site licensing authorities. Stakeholder input will be solicited from interested parties and incorporated during the development of the document. Currently only one consensus standard exists that explicitly deals with NDA holdup measurements: ASTM C1455 Standard Test Method for Nondestructive Assay of Special Nuclear Material Holdup Using Gamma-Ray Spectroscopic Methods. The ASTM International standard emphasizes the activities involved in actually making measurements, and was developed by safeguards and NDA experts. This new INMM ASC N15 standard will complement the existing ASTM international standard. One of the largest driving factors for writing this new standard was the recent emphasis on in situ NDA measurements by the safeguards community due to the Defense Nuclear Facility Safety Board (DNFSB) recommendation 2007-1 on in situ NDA measurements. Specifically, DNFSB recommendation 2007-1 referenced the lack of programmatic requirements for accurate in situ measurements and the use of measurement results for compliance with safety based requirements. That being the case, this paper will also discuss the progress made on the Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 2007-1 Safety-Related In Situ Nondestructive Assay of Radioactive Materials. Some of the information that will be presented includes observations made during site visits, how information useful to all facilities using nondestructive assay to determine holdup material quantities will be disseminated, and preliminary results of a gap analysis performed on current in situ nondestructive assay holdup measurements.

  9. CRAD, Facility Safety- Nuclear Facility Safety Basis

    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 Safety Basis.

  10. Nuclear Explosive Safety Study Process

    Energy Savers [EERE]

    Superseding DOE-STD-3015-97 January 1997 DOE STANDARD NUCLEAR EXPLOSIVE SAFETY STUDY PROCESS U.S. ... of high- consequence production, manufacturing, andor power plant operations. ...

  11. FAQS Reference Guide – Criticality Safety (NNSA)

    Broader source: Energy.gov [DOE]

    This reference guide has been developed to address the competency statements in DOE-STD-1173-2009, Criticality Safety Functional Area Qualification Standard.

  12. FAQS Reference Guide – Criticality Safety

    Broader source: Energy.gov [DOE]

    This reference guide addresses the competency statements in the April 2009 edition of DOE-STD-1173-2009, Criticality Safety Functional Area Qualification Standard.

  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. 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.

  15. Chief of Nuclear Safety | Department of Energy

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

    Chief of Nuclear Safety Chief of Nuclear Safety Message from Chief of Nuclear Safety Message from Chief of Nuclear Safety The Chief of Nuclear Safety (CNS) is responsible for ensuring that DOE Nuclear Safety Regulations, Standards, Guides, and national/international technical standards are applied in a correct manner in the conduct of DOE's nuclear mission under the purview of the Under Secretary for Management and Performance. Read more Seismic Lessons-Learned Panel Meetings Seismic

  16. Sandia Teaches Nuclear Safety Course

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

    Teaches Nuclear Safety Course - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs

  17. 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.

  18. 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.

  19. Office of Nuclear Safety and Environmental Assessments | Department of

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

    Energy Nuclear Safety and Environmental Assessments Office of Nuclear Safety and Environmental Assessments MISSION The Office of Nuclear Safety and Environmental Assessments conducts assessments to provide critical feedback and objective information on programs and performance in protecting our workers, the public and environment from radiological hazards with a focus on hazardous nuclear facilities and operations.This information provides assurance to our stakeholders and identifies areas

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

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

    Nuclear Safety Policy, Guidance & Reports Nuclear Safety Policy, Guidance & Reports The Office of Nuclear Safety establishes and maintains nuclear safety policy, requirements, and ...

  1. Office of Nuclear Safety Enforcement | Department of Energy

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

    Enforcement Office of Nuclear Safety Enforcement Office of Nuclear Safety Enforcement MISSION The Office of Nuclear Safety Enforcement implements the Department's nuclear safety ...

  2. Nuclear safety | Princeton Plasma Physics Lab

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

    safety Subscribe to RSS - Nuclear safety Actions taken to prevent nuclear and radiation ... NSTX Upgrade Like a new passenger jet or power plant, the National Spherical Torus ...

  3. Nuclear Safety Information | Department of Energy

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

    Nuclear Safety Information Nuclear Safety Information Idaho National Laboratory's Advanced Test Reactor (ATR) | April 8, 2009 Idaho National Laboratory's Advanced Test Reactor ...

  4. Nuclear Safety Reporting Criteria | Department of Energy

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

    Reporting Criteria Nuclear Safety Reporting Criteria January 1, 2012 Nuclear Safety Noncompliances Associated With Occurrences (DOE Order 232.2) These tables provide the criteria ...

  5. 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...

  6. 2012 Nuclear Safety Workshop Photos

    Broader source: Energy.gov [DOE]

    Deputy Secretary Poneman (view announcement memo) convened the second DOE Nuclear Safety Workshop on September 19-20, 2012. The event was held at the Bethesda North Marriott Hotel and Conference Center, 5701 Marinelli Road, Bethesda, MD.

  7. FUEL HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2005-06-30

    The purpose of this design calculation is to perform a criticality evaluation of the Fuel Handling Facility (FHF) and the operations and processes performed therein. The current intent of the FHF is to receive transportation casks whose contents will be unloaded and transferred to waste packages (WP) or MGR Specific Casks (MSC) in the fuel transfer bays. Further, the WPs will also be prepared in the FHF for transfer to the sub-surface facility (for disposal). The MSCs will be transferred to the Aging Facility for storage. The criticality evaluation of the FHF features the following: (I) Consider the types of waste to be received in the FHF as specified below: (1) Uncanistered commercial spent nuclear fuel (CSNF); (2) Canistered CSNF (with the exception of horizontal dual-purpose canister (DPC) and/or multi-purpose canisters (MPCs)); (3) Navy canistered SNF (long and short); (4) Department of Energy (DOE) canistered high-level waste (HLW); and (5) DOE canistered SNF (with the exception of MCOs). (II) Evaluate the criticality analyses previously performed for the existing Nuclear Regulatory Commission (NRC)-certified transportation casks (under 10 CFR 71) to be received in the FHF to ensure that these analyses address all FHF conditions including normal operations, and Category 1 and 2 event sequences. (III) Evaluate FHF criticality conditions resulting from various Category 1 and 2 event sequences. Note that there are currently no Category 1 and 2 event sequences identified for FHF. Consequently, potential hazards from a criticality point of view will be considered as identified in the ''Internal Hazards Analysis for License Application'' document (BSC 2004c, Section 6.6.4). (IV) Assess effects of potential moderator intrusion into the fuel transfer bay for defense in depth. The SNF/HLW waste transfer activity (i.e., assembly and canister transfer) that is being carried out in the FHF has been classified as safety category in the ''Q-list'' (BSC 2003, p. A-6). Therefore, this design calculation is subject to the requirements of the ''Quality Assurance Requirements and Description'' (DOE 2004), even though the FHF itself has not yet been classified in the Q-list. Performance of the work scope as described and development of the associated technical product conform to the procedure AP-3.124, ''Design Calculations and Analyses''.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. Review and Approval of Nuclear Facility Safety Basis and Safety...

    Office of Environmental Management (EM)

    DOE STANDARD REVIEW AND APPROVAL OF NUCLEAR FACILITY SAFETY BASIS AND SAFETY DESIGN BASIS ... Neither a reviewer nor the preparer has veto power over ultimate resolution or ...

  13. Nuclear Safety Regulatory Framework | Department of Energy

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

    Nuclear Safety Regulatory Framework Nuclear Safety Regulatory Framework February 2012 Presentation that outlines the rules, policies and orders that comprise the Department of Energy Nuclear Safety Regulatory Framework. PDF icon Nuclear Safety Regulatory Framework More Documents & Publications Summary Pamphlet, Nuclear Safety at the Department of Energy Notice of Violation, UChicago Argonne, LLC - WEA-2009-04 Independent Oversight Assessment, Waste Treatment and Immobilization Plant -

  14. 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; Guido, Joe

    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)

  15. 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.

  16. 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.

  17. 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.

  18. Nuclear Safety Specialist Functional Area Qualification Standard

    Energy Savers [EERE]

    83-2007 November 2007 DOE STANDARD NUCLEAR SAFETY SPECIALIST FUNCTIONAL AREA QUALIFICATION ... Center (NTC) * Institute of Nuclear Power Operations (INPO) * American Institute ...

  19. Princeton Plasma Physics Lab - Nuclear safety

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

    safety Actions taken to prevent nuclear and radiation accidents or to limit their consequences. en COLLOQUIUM: Technical Aspects of the Iran Nuclear Agreement http:www.pppl.gov...

  20. Nuclear Safety Research and Development Proposal Review and Prioritiza...

    Energy Savers [EERE]

    Nuclear Safety Research and Development Proposal Review and Prioritization Process and Criteria Nuclear Safety Research and Development Program Office of Nuclear Safety Office of ...

  1. Nuclear Safety Research and Development Committee Charter | Department...

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

    Committee Charter Nuclear Safety Research and Development Committee Charter July 5, 2012 Nuclear Safety Research and Development Committee Charter The intent of the Nuclear Safety ...

  2. Office of Nuclear Safety and Environmental Assessments | Department...

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

    Nuclear Safety and Environmental Assessments Office of Nuclear Safety and Environmental Assessments MISSION The Office of Nuclear Safety and Environmental Assessments conducts ...

  3. 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...

  4. 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.

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

    Energy Savers [EERE]

    Regulatory Commission, Office of Nuclear Material Safety and Safeguards, and the U.S. ... and nuclear material operations (DOE does not own or operate power reactors). ...

  6. Review and Approval of Nuclear Facility Safety Basis and Safety...

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

    104-2014, Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Documents by Website Administrator This Standard describes a framework and the criteria to be...

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

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

    5 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) provides objectives,...

  8. 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.

  9. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2005-04-26

    The purpose of this design calculation is to demonstrate that the handling operations of transportation casks performed in the Transportation Cask Receipt and Return Facility (TCRRF) and Buffer Area meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC [Bechtel SAIC Company] 2004 [DIRS 171599], Section 4.9.2.2), and the functional nuclear criticality safety requirement described in the ''Transportation Cask Receipt/Return Facility Description Document'' (BSC 2004 [DIRS 170217], Section 3.2.3). Specific scope of work contained in this activity consists of the following items: (1) Evaluate criticality effects for both dry and fully flooded conditions pertaining to TCRRF and Buffer Area operations for defense in depth. (2) Evaluate Category 1 and 2 event sequences for the TCRRF as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). This evaluation includes credible fuel reconfiguration conditions. In addition to the scope of work listed above, an evaluation was also performed of modeling assumptions for commercial spent nuclear fuel (CSNF) regarding inclusion of plenum and end regions of the active fuel. This calculation is limited to CSNF and US Department of Energy (DOE) SNF. it should be mentioned that the latter waste form is evaluated more in depth in the ''Canister Handling Facility Criticality Safety Calculations (BSC 2004 [DIRS 167614]). Further, the design and safety analyses of the naval SNF canisters are the responsibility of the US Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the TCRRF and Buffer Area and may not reflect the ongoing design evolution of the facility. However, it is anticipated that design changes to the facility layout will have little or no impact on the criticality results and/or conclusions presented in this document. This calculation is subject to the ''Quality Assurance Requirements and Description'' (DOE 2004 [DIRS 171539]) because the TCRRF is included in the Q-List (BSC 2004 [DIRS 168361], p. A-3) as an item important to safety. This calculation is prepared in accordance with AP-3.12Q, ''Design Calculations and Analyses'' [DIRS 168413].

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

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

    Criticality Safety Controls Implementation, May 31, 2013 (HSS CRAD 45-18, Rev. 1) Criticality Safety Controls Implementation, May 31, 2013 (HSS CRAD 45-18, Rev. 1) The Department ...

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

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

    Nuclear safety information sharing agreement between NRC and DOE's Office of Environment, Health, Safety and Security Nuclear safety information sharing agreement between NRC and ...

  12. 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.

  13. Nuclear Safety Software & Quality Assurance | Department of Energy

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

    Nuclear Safety Software & Quality Assurance In support of DOE O 410.1, Central Technical Authority Responsibilities Regarding Nuclear Safety Requirements, the Chief of Nuclear...

  14. Nuclear Explosives Safety Study Functional Area Qualification...

    Energy Savers [EERE]

    85-2007 September 2007 DOE STANDARD NUCLEAR EXPLOSIVE SAFETY STUDY FUNCTIONAL AREA QUALIFICATION ... bombs, including: * Radar fuses * Contact fuses * Timer fuses * Power supplies b. ...

  15. Preparation Of Nonreactor Nuclear Facility Documented Safety...

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

    9-2014, Preparation Of Nonreactor Nuclear Facility Documented Safety Analysis by Website Administrator This Department of Energy (DOE) Standard (STD), DOE-STD-3009-2014, describes...

  16. Quickstart Guide, Nuclear Safety Information Dashboard - September 2012 |

    Energy Savers [EERE]

    Department of Energy Quickstart Guide, Nuclear Safety Information Dashboard - September 2012 Quickstart Guide, Nuclear Safety Information Dashboard - September 2012 September 2012 Quickstart guide on how to use the features of Nuclear Safety Information dashboard tool. PDF icon Quickstart Guide, Nuclear Safety Information Dashboard - September 2012 More Documents & Publications Development of the Nuclear Safety Information Dashboard - September 2012 Responses to Frequently Asked

  17. nuclear safety | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response ...

  18. Neutron absorbing coating for nuclear criticality control

    DOE Patents [OSTI]

    Mizia, Ronald E.; Wright, Richard N.; Swank, William D.; Lister, Tedd E.; Pinhero, Patrick J.

    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.

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

    Energy Savers [EERE]

    Safety and Security HSS Independent Oversight Assessment of Nuclear Safety Culture and ... Commission (NRC), several nuclear power generating utilities, and associated ...

  20. Processing Exemptions to Nuclear Safety Rules and Approval of...

    Office of Environmental Management (EM)

    STD-1083-2009 June 2009 DOE STANDARD PROCESSING EXEMPTIONS TO NUCLEAR SAFETY RULES AND ... "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants." ...

  1. Processing Exemptions to Nuclear Safety Rules and Approval of...

    Office of Environmental Management (EM)

    DOE STANDARD PROCESSING EXEMPTIONS TO NUCLEAR SAFETY RULES AND APPROVAL OF ALTERNATIVE ... "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants." ...

  2. Nuclear Safety Research and Development Program Proposal Submittal...

    Energy Savers [EERE]

    5 Nuclear Safety Research and Development Program Proposal Submittal Instructions for Fiscal Year 2016 1.0 INTRODUCTION The Nuclear Safety Research and Development (NSR&D) Program ...

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

    Energy Savers [EERE]

    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 ...

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

    Office of Environmental Management (EM)

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

  5. Nuclear Safety Research and Development Program Operating Plan...

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

    Program Operating Plan Nuclear Safety Research and Development Program Operating Plan July 5, 2012 Nuclear Safety Research and Development Program Operating Plan This operating ...

  6. Nuclear Safety Research and Development Status Workshop Summary

    Office of Environmental Management (EM)

    Garzon 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...

  7. 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...

  8. 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.

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

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

    A Critical Step Toward Sustainable Nuclear Fuel Disposal A Critical Step Toward Sustainable Nuclear Fuel Disposal January 26, 2012 - 2:30pm Addthis Secretary Chu Secretary Chu...

  10. FAQS Reference Guide – Nuclear Safety Specialist

    Broader source: Energy.gov [DOE]

    This reference guide has been developed to address the competency statements in the November 2007 edition of DOE Standard DOE-STD-1183-2007, Nuclear Safety Specialist Functional Area Qualification Standard.

  11. Nuclear Safety Workshop Summary | Department of Energy

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

    On September 19-20, 2012, the U.S. Department of Energy (DOE) held a second Nuclear Safety Workshop covering the results of the Department's actions to improve its posture for...

  12. 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."

  13. Code of Federal Regulations NUCLEAR SAFETY MANAGEMENT

    Broader source: Energy.gov [DOE]

    This part governs the conduct of DOE contractors, DOE personnel, and other persons conducting activities (including providing items and services) that affect, or may affect, the safety of DOE nuclear facilities.

  14. 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...

  15. 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.

  16. 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).

  17. Nuclear Safety Research and Development Annual Report, December 2014 |

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

    Department of Energy Nuclear Safety Research and Development Annual Report, December 2014 Nuclear Safety Research and Development Annual Report, December 2014 December 8, 2014 - 1:22pm Addthis Nuclear Safety Research and Development Annual Report, December 2014 This document is the first annual report of DOE's Nuclear Safety Research and Development (NSR&D) Program, managed by the Office of Nuclear Safety in the Office of Environment, Health, Safety and Security. The report includes a

  18. Nuclear Safety Software & Quality Assurance | Department of Energy

    Energy Savers [EERE]

    Safety Software & Quality Assurance Nuclear Safety Software & Quality Assurance Nuclear Safety Software & Quality Assurance In support of DOE O 410.1, Central Technical Authority Responsibilities Regarding Nuclear Safety Requirements, the Chief of Nuclear Safety (CNS) provides operational awareness, oversight, and assistance to Environmental Management (EM) Headquarters, field offices, and their contractors in the areas of nuclear safety Quality Assurance (QA) and Software Quality

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

    Energy Savers [EERE]

    Message from Chief of Nuclear Safety Message from Chief of Nuclear Safety Message from Chief of Nuclear Safety The Chief of Nuclear Safety (CNS) is responsible for ensuring that DOE Nuclear Safety Regulations, Standards, Guides, and national/international technical standards are applied in a correct manner in the conduct of DOE's nuclear mission under the purview of the Under Secretary for Management and Performance. The Secretary of Energy committed the Department to having Central Technical

  20. safety | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure ...

  1. 2012 Nuclear Safety Workshop Presentations

    Broader source: Energy.gov [DOE]

    Lists workshop presentations from: Wednesday, September 19 - Plenary Session Wednesday, September 19 - Beyond Design Basis Events Analysis and Response Breakout Session Wednesday, September 19 - Safety Culture Breakout Session Wednesday, September 19 - Risk Assessment and Management Breakout Session Thursday, September 20 - Beyond Design Basis Events Analysis and Response Breakout Session Thursday, September 20 - Safety Culture Breakout Session Thursday, September 20 - Risk Assessment and Management Breakout Session Thursday, September 20 - Plenary Session

  2. Criticality Safety Validation of Scale 6.1

    SciTech Connect (OSTI)

    Marshall, William BJ J; Rearden, Bradley T

    2011-11-01

    The computational bias of criticality safety computer codes must be established through the validation of the codes to critical experiments. A large collection of suitable experiments has been vetted by the International Criticality Safety Benchmark Experiment Program (ICSBEP) and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). A total of more than 350 cases from this reference have been prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID) maintained by the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The performance of the KENO V.a and KENO-VI Monte Carlo codes within the Scale 6.1 code system with ENDF/B-VII.0 cross-section data in 238-group and continuous energy is assessed using the VALID models of benchmark experiments. The TSUNAMI tools for sensitivity and uncertainty analysis are utilized to examine some systems further in an attempt to identify potential causes of unexpected results. The critical experiments available for validation of the KENO V.a code cover eight different broad categories of systems. These systems use a range of fissile materials including a range of uranium enrichments, various plutonium isotopic vectors, and mixed uranium-plutonium oxides. The physical form of the fissile material also varies and is represented as metal, solutions, or arrays of rods or plates in a water moderator. The neutron energy spectra of the systems also vary and cover both fast and thermal spectra. Over 300 of the total cases used utilize the KENO V.a code. The critical experiments available for the validation of the KENO-VI code cover three broad categories of systems. The fissile materials in the systems vary and include high and intermediate-enrichment uranium and mixed uranium/plutonium oxides. The physical form of the fissile material is either metal or rod arrays in water. As with KENO V.a, both fast and thermal neutron energy spectra are represented in the systems considered. The results indicate generally good performance of both the KENO V.a and KENO-VI codes across the range of systems analyzed. The bias of calculated k{sub eff} from expected values is less than 0.9% {Delta}k in all cases. All eight categories of experiments show biases of less than 0.5% {Delta}k in KENO V.a with the exception of intermediate enrichment metal systems using the 238-group library. The continuous energy library generally manifests lower biases than the multi-group data. The KENO-VI results show slightly larger biases, though this may primarily be the result of modeling systems with more geometric complexity, which are more difficult to describe accurately, even with a generalized geometry code like KENO-VI. Several additional conclusions can be drawn from the results of this validation effort. These conclusions include that the TSUNAMI tools can be used successfully to explain the cause of aberrant results, that some evaluations in the IHECSBE should be updated to provide more rigorous expected k{sub eff} values and uncertainties, and that potential cross-section errors can be identified by detailed review of the results of this validation. It also appears that the overall cross-section uncertainty as quantified through the Scale covariance library is overestimated. Overall, the KENO V.a and KENO-VI codes are shown to provide consistent, low bias results for a wide range of physical systems of potential interest in criticality safety applications.

  3. 2016 Nuclear and Facility Safety Program Workshop | Department of Energy

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

    Nuclear and Facility Safety Program Workshop 2016 Nuclear and Facility Safety Program Workshop March 22, 2016 - 3:48pm Addthis 2016 Nuclear and Facility Safety Program Workshop The Office of Environmental Health, Safety, and Security will sponsor the 2016 Nuclear and Facility Safety Program Workshop which will be held May 2-6, 2016 at the Alexis Park in Las Vegas, Nevada. The Workshop will include meetings for the DOE Safety Culture Improvement Panel, Federal Technical Capability Panel, Facility

  4. 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.

  5. 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.

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

    SciTech Connect (OSTI)

    Marshall, William BJ J; Rearden, Bradley T

    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.

  7. safety and security | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    safety and security

  8. Improved Nuclear Safety Through International Standards

    SciTech Connect (OSTI)

    Doctor, Steven R.; Moffitt, Robert L.; Taylor, Theodore T.; Trosman, Grigory

    1999-12-01

    This paper describes the 1986 Chornobyl accident, notes some of its effects, and reviews the cause. International efforts to improve reactor safety to prevent another such accident are listed. The U.S. Department of Energy (DOE) program to improve the safety of Soviet-designed nuclear power plants is outlined, followed by a more detailed description of the specific projects related to nondestructive evaluation. Future directions are proposed, and conclusions are provided.

  9. 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.

  10. 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.

  11. Nuclear Safety Research and Development Annual Report, December 2014 |

    Energy Savers [EERE]

    Reporting Criteria Nuclear Safety Reporting Criteria January 1, 2012 Nuclear Safety Noncompliances Associated With Occurrences (DOE Order 232.2) These tables provide the criteria for reporting nuclear safety noncompliances into the Department of Energy's Noncompliance Tracking System (NTS). A more detailed description of the NTS reporting criteria and expectations can be found in the Office of Health, Safety and Security's Enforcement Coordinator Handbook. PDF icon Nuclear Safety Reporting

  12. 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

  13. Chief of Nuclear Safety (CNS) Staff Assignments & Expertise ...

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

    Chief of Nuclear Safety (CNS) Staff Assignments & Expertise CNS staff maintains adequate technical proficiency, including the timely completion of Senior Technical Safety Manager...

  14. 2015 Nuclear & Facility Safety Programs Workshop Agenda | Department...

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

    2015 Nuclear and Facility Safety Programs Workshop agenda outlining following: Training Plenary Session Award Presentations Guest speakers Fire Safety Workshop Facility...

  15. Nuclear safety information sharing agreement between NRC and DOE's Office

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

    of Environment, Health, Safety and Security | Department of Energy Nuclear safety information sharing agreement between NRC and DOE's Office of Environment, Health, Safety and Security Nuclear safety information sharing agreement between NRC and DOE's Office of Environment, Health, Safety and Security December 2014 agreement between NRC and DOE's Office of Environment, Health, Safety and Security to establish a framework for DOE and NRC to exchange information related to safety issues

  16. 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).

  17. Assessment of Nuclear Safety Culture at the Salt Waste Processing...

    Office of Environmental Management (EM)

    Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project ... Commission (NRC), several nuclear power generating utilities, and associated ...

  18. Assessment of Nuclear Safety Culture at the Idaho Cleanup Project...

    Office of Environmental Management (EM)

    Independent Oversight Assessment of Nuclear Safety Culture at the Idaho Cleanup Project ... Commission (NRC), several nuclear power generating utilities, and associated ...

  19. 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...

  20. Microsoft Word - Nuclear Safety Pamphlet Final September 1 2010...

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

    A Basic Overview of NUCLEAR SAFETY AT THE DEPARTMENT OF ENERGY Outreach & Awareness Series ... (DOE) is committed to conducting its nuclear operations in a manner that protects the ...

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

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

    of Energy's (DOE) nuclear facilities and identify opportunities for improvement. Nuclear Safety Workshop Agenda - Post Fukushima Initiatives and Results More Documents &...

  2. Summary Pamphlet, Nuclear Safety at the Department of Energy

    Broader source: Energy.gov [DOE]

    This pamphlet is intended to provide an abbreviated summary of regulatory requirements and processes for ensuring nuclear safety at DOE, which serve as the Department’s overarching regulatory framework for nuclear safety.

  3. Deputy Secretary Poneman Discusses Nuclear Safety at the IAEA...

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

    Discusses Nuclear Safety at the IAEA Deputy Secretary Poneman Discusses Nuclear Safety at the IAEA June 20, 2011 - 12:00am Addthis Washington, D.C. - U.S. Deputy Secretary of ...

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

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

    Cites Fluor Fernald Inc. for Nuclear Safety Violations DOE Cites Fluor Fernald Inc. for Nuclear Safety Violations August 25, 2005 - 2:43pm Addthis Washington, D.C. - The Department...

  5. CRAD, Nuclear Facility Safety System- September 25, 2009

    Broader source: Energy.gov [DOE]

    Nuclear Facility Safety System Functionality Inspection Criteria, Inspection Activities, and Lines of Inquiry (HSS CRAD 64-17, Rev 0 )

  6. Defense Nuclear Facilities Safety Board (DNFSB) Letters and Recommendations

    Energy Savers [EERE]

    | Department of Energy Defense Nuclear Facilities Safety Board (DNFSB) Letters and Recommendations Defense Nuclear Facilities Safety Board (DNFSB) Letters and Recommendations Defense Nuclear Facilities Safety Board (DNFSB) Letters and Recommendations The Defense Nuclear Facilities Safety Board (DNFSB) is an independent organization within the executive branch chartered with the responsibility of providing recommendations and advice to the President and the Secretary of Energy regarding

  7. 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.

  8. 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.

  9. Double-clad nuclear fuel safety rod

    DOE Patents [OSTI]

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

    1984-01-01

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

  10. 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.

  11. DOE Cites Safety and Ecology Corp. for Violating Nuclear Safety Rules |

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

    Department of Energy 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 WASHINGTON, D.C. -- The Department of Energy (DOE) today notified Safety and Ecology Corporation, the contractor responsible for radiological safety at the Portsmouth Gaseous Diffusion Project in Portsmouth, Ohio, that it will fine the company $55,000 for violating the department's regulations prohibiting

  12. 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).

  13. 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.

  14. 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.

  15. 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.

    2014-11-01

    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. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.

  16. 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.

  17. Approach for validating actinide and fission product compositions for burnup credit criticality safety analyses

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

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

    2014-11-01

    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 ofmore » 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. Furthermore, the validation approach and results support a recent revision of the U.S. Nuclear Regulatory Commission Interim Staff Guidance 8.« less

  18. Chief of Nuclear Safety (CNS) Staff Assignments & Expertise

    Broader source: Energy.gov [DOE]

    CNS staff maintains adequate technical proficiency, including the timely completion of Senior Technical Safety Manager (STSM) qualification. Further, CNS staff periodically review and assess whether EM is maintaining adequate numbers of technically competent personnel necessary to fulfill its nuclear safety responsibilities and identify gaps in nuclear safety-related technical human capital.

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

    Broader source: Energy.gov [DOE]

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

  20. CRAD, Nuclear Safety Delegations for Documented Safety Analysis Approval – January 8, 2015 (EA CRAD 31-09, Rev. 0)

    Broader source: Energy.gov [DOE]

    CRAD, Nuclear Safety Delegations for Documented Safety Analysis Approval – January 8, 2015 (EA CRAD 31-09, Rev. 0)

  1. 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.

  2. Propagation of Isotopic Bias and Uncertainty to Criticality Safety Analyses of PWR Waste Packages

    SciTech Connect (OSTI)

    Radulescu, Georgeta

    2010-06-01

    Burnup credit methodology is economically advantageous because significantly higher loading capacity may be achieved for spent nuclear fuel (SNF) casks based on this methodology as compared to the loading capacity based on a fresh fuel assumption. However, the criticality safety analysis for establishing the loading curve based on burnup credit becomes increasingly complex as more parameters accounting for spent fuel isotopic compositions are introduced to the safety analysis. The safety analysis requires validation of both depletion and criticality calculation methods. Validation of a neutronic-depletion code consists of quantifying the bias and the uncertainty associated with the bias in predicted SNF compositions caused by cross-section data uncertainty and by approximations in the calculational method. The validation is based on comparison between radiochemical assay (RCA) data and calculated isotopic concentrations for fuel samples representative of SNF inventory. The criticality analysis methodology for commercial SNF disposal allows burnup credit for 14 actinides and 15 fission product isotopes in SNF compositions. The neutronic-depletion method for disposal criticality analysis employing burnup credit is the two-dimensional (2-D) depletion sequence TRITON (Transport Rigor Implemented with Time-dependent Operation for Neutronic depletion)/NEWT (New ESC-based Weighting Transport code) and the 44GROUPNDF5 crosssection library in the Standardized Computer Analysis for Licensing Evaluation (SCALE 5.1) code system. The SCALE 44GROUPNDF5 cross section library is based on the Evaluated Nuclear Data File/B Version V (ENDF/B-V) library. The criticality calculation code for disposal criticality analysis employing burnup credit is General Monte Carlo N-Particle (MCNP) Transport Code. The purpose of this calculation report is to determine the bias on the calculated effective neutron multiplication factor, k{sub eff}, due to the bias and bias uncertainty associated with predicted spent fuel compositions (i.e., determine the penalty in reactivity due to isotopic composition bias and uncertainty) for use in disposal criticality analysis employing burnup credit. The method used in this calculation to propagate the isotopic bias and bias-uncertainty values to k{sub eff} is the Monte Carlo uncertainty sampling method. The development of this report is consistent with 'Test Plan for: Isotopic Validation for Postclosure Criticality of Commercial Spent Nuclear Fuel'. This calculation report has been developed in support of burnup credit activities for the proposed repository at Yucca Mountain, Nevada, and provides a methodology that can be applied to other criticality safety applications employing burnup credit.

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

    SciTech Connect (OSTI)

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

    2013-07-01

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

  4. DOE Cites Bechtel Jacobs Company for Nuclear Safety Violations | Department

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

    of Energy Jacobs Company for Nuclear Safety Violations DOE Cites Bechtel Jacobs Company for Nuclear Safety Violations August 4, 2005 - 2:36pm Addthis WASHINGTON, D.C. - The Department of Energy (DOE) today notified the Bechtel Jacobs Company (BJC) that it will fine the company $247,500 for violations of the department's nuclear safety requirements. The company is the department's contractor responsible for environmental cleanup and waste management at its Oak Ridge Reservation in Tennessee.

  5. DOE Cites Washington TRU Solutions for Nuclear Safety Violations |

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

    Department of Energy TRU Solutions for Nuclear Safety Violations DOE Cites Washington TRU Solutions for Nuclear Safety Violations December 22, 2005 - 4:53pm Addthis WASHINGTON, D.C. -- The Department of Energy (DOE) today notified Washington TRU Solutions (WTS) that it will fine the company $192,500 for violations of the department's nuclear safety requirements. The Preliminary Notice of Violation (PNOV) issued today cites a number of deficiencies that led to a series of low-level plutonium

  6. Defense Nuclear Facilities Safety Board (DNFSB) Update - Dale Govan,

    Energy Savers [EERE]

    Departmental Representative to the DNFSB | Department of Energy Defense Nuclear Facilities Safety Board (DNFSB) Update - Dale Govan, Departmental Representative to the DNFSB Defense Nuclear Facilities Safety Board (DNFSB) Update - Dale Govan, Departmental Representative to the DNFSB DNFSB Mission The Board provides independent analysis, advice and recommendations to the Secretary to ensure adequate protection to public health and safety at defense nuclear facilities. Identify Department

  7. Sandia Nuclear Power Safety Expert Elected to National Academy...

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

    Nuclear Power Safety Expert Elected to National Academy of Engineering - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure ...

  8. NNSA and Defense Nuclear Facilities Safety Board certifications...

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

    allocated funding NNSA and Defense Nuclear Facilities Safety Board certifications free up 47 million in previously allocated funding The DNFSB and NNSA required the CMRR...

  9. 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.

  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 DNFSB Mission The Board provides independent analysis, advice and ...

  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. 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.

  13. Senior Technical Safety Manager Qualification Program Self-Assessment- Chief of Nuclear Safety

    Broader source: Energy.gov [DOE]

    This Chief of Nuclear Safety (CNS) Report was prepared to summarize the results of the July 2013 CNS self-assessment of the Senior Technical Safety Manager Qualification Program.

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

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

    Energy Nuclear Safety Enforcement Regulations and Directives - Nuclear Safety 10 C.F.R. Part 820 and Amendments 10 C.F.R. Part 820 - Procedural Rules for DOE Nuclear Activities 10 C.F.R. Part 820 - Procedural Rules for DOE Nuclear Activities; General Statement of Enforcement Policy; Final rule; amendment of enforcement policy statement and confirmation of interim rule Enforcement Guidance Supplements 10 C.F.R. Part 830 10 C.F.R. Part 830 - Nuclear Safety Management; Final Rule Office of

  15. Operating Experience Level 3, Importance of Conduct of Operations and Training for Effective Criticality Safety Programs

    Broader source: Energy.gov [DOE]

    OE-3 2012-07: Importance of Conduct of Operations and Training for Effective Criticality Safety Programs

  16. 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

  17. 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.

  18. 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.

  19. Criticality safety evaluation report for FFTF 42% fuel assemblies

    SciTech Connect (OSTI)

    Richard, R.F.

    1997-10-28

    An FFTF tritium/isotope production mission will require a new fuel supply. The reference design core will use a mixed oxide fuel nominally enriched to 40 wt% Pu. This enrichment is significantly higher than that of the standard Driver Fuel Assemblies used in past operations. Consequently, criticality safety for handling and storage of this fuel must be addressed. The purpose of this document is to begin the process by determining the minimum critical number for these new fuel assemblies in water, sodium and air. This analysis is preliminary and further work can be done to refine the results reported here. Analysis was initially done using 45 wt 5 PuO. Additionally, a preliminary assessment is done concerning storage of these fuel assemblies in Interim Decay Storage (IDS), Fuel Storage Facility (FSF), and Core Component Containers/Interim Storage Casks (CCC/ISC).

  20. A Critical Step Toward Sustainable Nuclear Fuel Disposal | Department of

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

    Energy A Critical Step Toward Sustainable Nuclear Fuel Disposal A Critical Step Toward Sustainable Nuclear Fuel Disposal January 26, 2012 - 2:30pm Addthis Secretary Chu Secretary Chu Former Secretary of Energy The Blue Ribbon Commission on America's Nuclear Future was formed at the direction of the President to conduct a comprehensive review of polices for managing the back end of the nuclear fuel cycle. If we are going to ensure that the United States remains at the forefront of nuclear

  1. Facility Safety - DOE Directives, Delegations, and Requirements

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

    facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH)...

  2. Packaging Strategies for Criticality Safety for "Other" DOE Fuels in a Repository

    SciTech Connect (OSTI)

    Larry L Taylor

    2004-06-01

    Since 1998, there has been an ongoing effort to gain acceptance of U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in the national repository. To accomplish this goal, the fuel matrix was used as a discriminating feature to segregate fuels into nine distinct groups. From each of those groups, a characteristic fuel was selected and analyzed for criticality safety based on a proposed packaging strategy. This report identifies and quantifies the important criticality parameters for the canisterized fuels within each criticality group to: (1) demonstrate how the other fuels in the group are bounded by the baseline calculations or (2) allow identification of individual type fuels that might require special analysis and packaging.

  3. Criticality safety evaluation report for K Basin filter cartridges

    SciTech Connect (OSTI)

    Schwinkendorf, K.N.

    1995-01-01

    A criticality safety evaluation of the K Basin filter cartridge assemblies has been completed to support operations without a criticality alarm system. The results show that for normal operation, the filter cartridge assembly is far below the safety limit of k{sub eff} = 0.95, which is applied to plutonium systems at the Hanford Site. During normal operating conditions, uranium, plutonium, and fission and corrosion products in solution are continually accumulating in the available void spaces inside the filter cartridge medium. Currently, filter cartridge assemblies are scheduled to be replaced at six month intervals in KE Basin, and at one year intervals in KW Basin. According to available plutonium concentration data for KE Basin and data for the U/Pu ratio, it will take many times the six-month replacement time for sufficient fissionable material accumulation to take place to exceed the safety limit of k{sub eff} = 0.95, especially given the conservative assumption that the presence of fission and corrosion products is ignored. Accumulation of sludge with a composition typical of that measured in the sand filter backwash pit will not lead to a k{sub eff} = 0.95 value. For off-normal scenarios, it would require at least two unlikely, independent, and concurrent events to take place before the k{sub eff} = 0.95 limit was exceeded. Contingencies considered include failure to replace the filter cartridge assemblies at the scheduled time resulting in additional buildup of fissionable material, the loss of geometry control from the filter cartridge assembly breaking apart and releasing the individual filter cartridges into an optimal configuration, and concentrations of plutonium at U/Pu ratios less than measured data for KE Basin, typically close to 400 according to extensive measurements in the sand filter backwash pit and plutonium production information.

  4. Submersion criticality safety of tungsten-rhenium urania cermet fuel for space propulsion and power applications

    SciTech Connect (OSTI)

    A.E. Craft; R. C. O'Brien; S. D. Howe; J. C. King

    2014-07-01

    Nuclear thermal rockets are the preferred propulsion technology for a manned mission to Mars, and tungsten–uranium oxide cermet fuels could provide significant performance and cost advantages for nuclear thermal rockets. A nuclear reactor intended for use in space must remain subcritical before and during launch, and must remain subcritical in launch abort scenarios where the reactor falls back to Earth and becomes submerged in terrestrial materials (including seawater, wet sand, or dry sand). Submersion increases reflection of neutrons and also thermalizes the neutron spectrum, which typically increases the reactivity of the core. This effect is typically very significant for compact, fast-spectrum reactors. This paper provides a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor with a range of fuel compositions. Each submersion case considers both the rhenium content in the matrix alloy and the uranium oxide volume fraction in the cermet. The inclusion of rhenium significantly improves the submersion criticality safety of the reactor. While increased uranium oxide content increases the reactivity of the core, it does not significantly affect the submersion behavior of the reactor. There is no significant difference in submersion behavior between reactors with rhenium distributed within the cermet matrix and reactors with a rhenium clad in the coolant channels. The combination of the flooding of the coolant channels in submersion scenarios and the presence of a significant amount of spectral shift absorbers (i.e. high rhenium concentration) further decreases reactivity for short reactor cores compared to longer cores.

  5. Exploratory Nuclear Reactor Safety Analysis and Visualization via

    Office of Scientific and Technical Information (OSTI)

    Integrated Topological and Geometric Techniques (Technical Report) | SciTech Connect Technical Report: Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques Citation Details In-Document Search Title: Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming

  6. 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.

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

    Broader source: Energy.gov [DOE]

    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.

  8. 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.

  9. Nuclear Safety Research and Development (NSR&D) Program | Department of

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

    Energy Safety Research and Development (NSR&D) Program Nuclear Safety Research and Development (NSR&D) Program The Nuclear Safety Research and Development (NSR&D) Program is managed by the Office of Nuclear Safety, within the Office of Environment, Health, Safety and Security (AU) to provide corporate-level leadership supporting nuclear safety research and development throughout the Department of Energy (DOE). The NSR&D Program also consults with the Nuclear Safety Council,

  10. Office of Safety | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Gallery Photo Gallery Jobs Apply for Our Jobs Our Jobs Working at NNSA Blog Home Office of Safety Office of Safety NNSA's Asset Management Program Completes First Pilot The...

  11. Nuclear Safety Enforcement Documents | Department of Energy

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

    Exposure at the Hanford Site June 14, 2005 Preliminary Notice of Violation, Safety and Ecology Corporation - EA-2005-03 Issued to Safety and Ecology Corporation related to a 10 CFR...

  12. An advanced deterministic method for spent fuel criticality safety analysis

    SciTech Connect (OSTI)

    DeHart, M.D.

    1998-01-01

    Over the past two decades, criticality safety analysts have come to rely to a large extent on Monte Carlo methods for criticality calculations. Monte Carlo has become popular because of its capability to model complex, non-orthogonal configurations or fissile materials, typical of real world problems. Over the last few years, however, interest in determinist transport methods has been revived, due shortcomings in the stochastic nature of Monte Carlo approaches for certain types of analyses. Specifically, deterministic methods are superior to stochastic methods for calculations requiring accurate neutron density distributions or differential fluxes. Although Monte Carlo methods are well suited for eigenvalue calculations, they lack the localized detail necessary to assess uncertainties and sensitivities important in determining a range of applicability. Monte Carlo methods are also inefficient as a transport solution for multiple pin depletion methods. Discrete ordinates methods have long been recognized as one of the most rigorous and accurate approximations used to solve the transport equation. However, until recently, geometric constraints in finite differencing schemes have made discrete ordinates methods impractical for non-orthogonal configurations such as reactor fuel assemblies. The development of an extended step characteristic (ESC) technique removes the grid structure limitations of traditional discrete ordinates methods. The NEWT computer code, a discrete ordinates code built upon the ESC formalism, is being developed as part of the SCALE code system. This paper will demonstrate the power, versatility, and applicability of NEWT as a state-of-the-art solution for current computational needs.

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

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

    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. 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.

  15. Development of the Nuclear Safety Information Dashboard - September...

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

    A working group with nuclear safety expertise used paired pairing computer software to ... A computer program was used to combine the results for each "paired pair" in the group and ...

  16. Review and Approval of Nuclear Facility Safety Basis Documents...

    Energy Savers [EERE]

    CHANGE NOTICE NO. 3 Date December 2005 DOE STANDARD REVIEW AND APPROVAL OF NUCLEAR FACILITY SAFETY ... Neither a reviewer nor the preparer has veto power over ultimate resolution or ...

  17. Exelon Statement Regarding Nuclear Safety and 10 CFR 810

    Office of Energy Efficiency and Renewable Energy (EERE)

    Exelon respectfully submits that the existing 810 rule, as currently interpreted, and the proposed revised rule, both work as deterrents to improving safety in nuclear operations around the world.

  18. Fiscal Year 2016 Call for Nuclear Safety Research and Development...

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

    FROM: SUBJECT: Fiscal Year 2016 Call for Nuclear Safety Research and Development Proposals The purpose of this memorandum is to inform you of the Fiscal Year 2016 Call for ...

  19. 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.

  20. 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.

  1. 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.

  2. Spent nuclear fuel project path forward preliminary safety evaluation

    SciTech Connect (OSTI)

    Brehm, J.R.; Crowe, R.D.; Siemer, J.M.; Wojdac, L.F.; Hosler, A.G.

    1995-03-01

    This preliminary safety evaluation (PSE) provides validation of the initial project design criteria for the Spent Nuclear Fuel Project (SNFP) Path Forward for removal of fuel from K Basins.

  3. Sandia Nuclear Power Safety Expert Elected to National Academy of

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

    Engineering Nuclear Power Safety Expert Elected to National Academy of Engineering - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing

  4. 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.

  5. Nuclear and Facility Safety Directives | Department of Energy

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

    Directives Nuclear and Facility Safety Directives DOE Order (O) 252.1A, Technical Standards Program DOE O 252.1A promotes DOE's use of Voluntary Consensus Standards (VCS) as the primary method for application of technical standards and establishes and manages the DOE Technical Standards Program (TSP) including technical standards development, information, activities, issues, and interactions. AU-30 Contact: Jeff Feit DOE Policy (P) 420.1, Department of Energy Nuclear Safety Policy DOE P 420.1,

  6. 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.

  7. Supplemnental Volume - Independent Oversight Assessment of the Nuclear Safety Culture and Management of Nuclear Safety Concerns at the Hanford Site Waste Treatment and Immobilization Plant, January 2012

    Energy Savers [EERE]

    Supplemental Volume Independent Oversight Assessment of Nuclear Safety Culture and Management of Nuclear Safety Concerns at the Hanford Site Waste Treatment and Immobilization Plant January 2012 Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Office of Health, Safety and Security HSS i Independent Oversight Assessment of Safety Culture and Management of Nuclear Safety Concerns at the Hanford Site Waste Treatment and Immobilization Plant

  8. Nuclear Safety Research and Development Proposal Review and Prioritization and Criteria

    Energy Savers [EERE]

    Nuclear Safety Research and Development Proposal Review and Prioritization Process and Criteria Nuclear Safety Research and Development Program Office of Nuclear Safety Office of Environment, Health, Safety, and Security U.S. Department of Energy December 2015 NSR&D Proposal Review and Prioritization Process and Criteria 2 | P a g e Nuclear Safety Research and Development Program 1. Introduction Per the Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program

  9. FAQS Qualification Card – Nuclear Safety Specialist

    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).

  10. Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

    SciTech Connect (OSTI)

    Richard, R.F.

    1996-10-09

    This Criticality Safety Evaluation allows a mix of up to five pin containers plus two assemblies in the same Core Component Container.

  11. 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.

  12. 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.

  13. 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.

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

    SciTech Connect (OSTI)

    Slate, L.J.; Taylor, J.T.

    2000-08-31

    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.

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

    SciTech Connect (OSTI)

    Morris, Tommy J.

    2012-07-05

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

  16. 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.

  17. Nuclear Safety Specialist FTCP Functional Area Qualification...

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

    ... Discuss the application of the Center for Chemical Process Safety's Guidelines for Hazard ... DOE-STD-1104-2014. Objective B.: Perform reviews and determine the adequacy of the ...

  18. LANL Nuclear Safety Support Services IDIQ Contract

    Broader source: Energy.gov [DOE]

    The scope of this contract will consist of technical support services needed to support the development and implementation of Documented Safety Analysis for EM operations facilities at LANL, such as Technical Area 54 Area G.

  19. 2015 Nuclear & Facility Safety Programs Workshop

    Office of Energy Efficiency and Renewable Energy (EERE)

    The workshop will feature training opportunities, sharing of best practices and lessons-learned, thought-provoking discussions, and an award ceremony recognizing outstanding performance by DOE safety professionals.

  20. 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.

  1. 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.

  2. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    SciTech Connect (OSTI)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  3. Department of Energy Cites Nuclear Waste Partnership, LLC and Los Alamos National Security, LLC for Violations Related to Worker Safety and Health and Nuclear Safety

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) today issued a Preliminary Notice of Violation (PNOV) to Nuclear Waste Partnership, LLC (NWP) for violations of DOE worker safety and health and nuclear safety requirements.

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

    SciTech Connect (OSTI)

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

    2015-07-06

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

  5. 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

  6. 10 CFR Part 830 Nuclear Safety Technical Positions | Department of Energy

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

    10 CFR Part 830 Nuclear Safety Technical Positions 10 CFR Part 830 Nuclear Safety Technical Positions 10 CFR Part 830 Nuclear Safety Technical Positions Nuclear and Facility Safety Policy is the Office of Primary Interest (OPI) responsible for the development, interpretation, and revision of a number of DOE directives. Technical Positions to directives issued by Nuclear and Facility Safety Policy provide clarification for specific applications of the requirements in DOE orders, rules, and other

  7. A Look Back at the Nuclear Safety Workshop | Department of Energy

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

    Nuclear Safety Workshop A Look Back at the Nuclear Safety Workshop June 16, 2011 - 2:59pm Addthis Glenn Podonsky Glenn Podonsky Director, Independent Enterprise Assessments As the Department's Chief Health, Safety and Security Officer, my job is to make sure that we continue to enhance and improve the safety of the Energy Department's nuclear facilities. That is why, in response to the March accident at the Fukushima Daiichi nuclear complex, the Department hosted a Nuclear Safety Workshop to

  8. 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.

  9. safety

    National Nuclear Security Administration (NNSA)

    contractor at the Nevada National Security Site, has been recognized by the Department of Energy for excellence in occupational safety and health protection. National Nuclear...

  10. Measurement of Fresh Fuel Rods to Demonstrate Compliance with Criticality Safety Limits

    SciTech Connect (OSTI)

    Miko, David K.; Desimone, David J.

    2015-11-03

    In order to operate TA-66 as a radiological facility with the quantity of nuclear material required to fulfil its mission, a criticality safety evaluation was required. This evaluation defined the control parameters for operations at the facility. The resulting evaluation for TA-66 placed limits on the amount of SNM, as well as other materials such as beryllium. In addition, there is a limit on the number of uranium fuel rods allowed subject to enrichment, outer diameter, and overall length restrictions. The enrichments for the rods to be shipped to TA-66 were documented in LA-UR-13-23581, but the outer diameter and length were not documented. This report provides this information.

  11. 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.

  12. 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. 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, supersedes DOE O 420.1C.

  13. Nuclear and Facility Safety Policy Rules | Department of Energy

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

    Policy Rules Nuclear and Facility Safety Policy Rules DOE provides safety requirements and guidance in a number of forms. One form in which we publish requirements is through rulemaking. Federal rules and regulations are published in the Code of Federal Regulations (CFR) and are noticed for review and comment by members of the public in the Federal Register (FR) consistent with the Administrative Procedures Act. Requirements in rules apply to our contractors whether or not they are also

  14. 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.

  15. Nuclear Explosive Safety Study Functional Area Qualification Standard

    Energy Savers [EERE]

    i NOT MEASUREMENT SENSITIVE DOE-STD-1185-2007 CHANGE NOTICE No.1 April 2010 DOE STANDARD NUCLEAR EXPLOSIVE SAFETY STUDY FUNCTIONAL AREA QUALIFICATION STANDARD DOE Defense Nuclear Facilities Technical Personnel U.S. Department of Energy AREA TRNG Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. DOE-STD-1185-2007 ii This document is available on the Department of Energy Technical Standards Program Web Site at

  16. DOE Representative to World Institute of Nuclear Safety (WINS) | National

    National Nuclear Security Administration (NNSA)

    Nuclear Security Administration DOE Representative to World Institute of Nuclear Safety (WINS) Lisa G. Hilliard Lisa Hilliard August 2009 NNSA Administrator's Silver Award Lisa G. Hilliard has received the NNSA Administrator's Silver Award for her sustained distinguished accomplishments as the Office Director of the DOE office to the U.S. Mission to International Organizations in Vienna from May 1993 to April 2009, serving four Ambassadors, two interim Representatives, and six Secretaries of

  17. CRITICALITY SAFETY CONTROL OF LEGACY FUEL FOUND AT 105-K WEST FUEL STORAGE BASIN

    SciTech Connect (OSTI)

    JENSEN, M.A.

    2005-08-19

    In August 2004, two sealed canisters containing spent nuclear fuel were opened for processing at the Hanford Site's K West fuel storage basin. The fuel was to be processed through cleaning and sorting stations, repackaged into special baskets, placed into a cask, and removed from the basin for further processing and eventual dry storage. The canisters were expected to contain fuel from the old Hanford C Reactor, a graphite-moderated reactor fueled by very low-enriched uranium metal. The expected fuel type was an aluminum-clad slug about eight inches in length and with a weight of about eight pounds. Instead of the expected fuel, the two canisters contained several pieces of thin tubes, some with wire wraps. The material was placed into unsealed canisters for storage and to await further evaluation. Videotapes and still photographs of the items were examined in consultation with available retired Hanford employees. It was determined that the items had a fair probability of being cut-up pieces of fuel rods from the retired Hanford Plutonium Recycle Test Reactor (PRTR). Because the items had been safely handled several times, it was apparent that a criticality safety hazard did not exist when handling the material by itself, but it was necessary to determine if a hazard existed when combining the material with other known types of spent nuclear fuel. Because the PRTR operated more than 40 years ago, investigators had to rely on a combination of researching archived documents, and utilizing common-sense estimates coupled with bounding assumptions, to determine that the fuel items could be handled safely with other spent nuclear fuel in the storage basin. As older DOE facilities across the nation are shut down and cleaned out, the potential for more discoveries of this nature is increasing. As in this case, it is likely that only incomplete records will exist and that it will be increasingly difficult to immediately characterize the nature of the suspect fissionable material and its criticality hazards.

  18. Criticality Safety Controls Implementation, May 31, 2013 (HSS CRAD 45-18, Rev. 1)

    Broader source: Energy.gov [DOE]

    The Department of Energy (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 (see HSS CRAD 64-32). The following provides a set of criteria and typical activities with 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 management.

  19. Nuclear Safety Research and Development Program Proposal Submittal Instrustions for Fiscal Year 2016

    Energy Savers [EERE]

    Energy Committee Charter Nuclear Safety Research and Development Committee Charter July 5, 2012 Nuclear Safety Research and Development Committee Charter The intent of the Nuclear Safety Research and Development (NSR&D) Committee is to identify nuclear safety research needs and opportunities within the Department of Energy (DOE) and National Nuclear Security Administration (NNSA) and their program offices. The Committee promotes communication and coordination among DOE and NNSA program

  20. Development of the Nuclear Safety Information Dashboard- September 2012

    Broader source: Energy.gov [DOE]

    A working group with nuclear safety expertise used paired pairing computer software to develop first, a severity-weighted factor for the 17 Groups of ORPS Reporting Criteria and then, a severity-weighted factor for the sixty-five ORPS reporting criteria.

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

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

    Freels, James D.

    2016-01-01

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

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

    SciTech Connect (OSTI)

    Garvin, L.J.

    1995-11-01

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

  3. Criticality Safety Basics for INL FMHs and CSOs (Technical Report...

    Office of Scientific and Technical Information (OSTI)

    A paper copy of this document is also available for sale to the public from the National Technical Information Service, Springfield, VA at www.ntis.gov. Nuclear power is a valuable ...

  4. NSS 18.1 Criticality Safety 5/26/95

    Broader source: Energy.gov [DOE]

    The objective of this surveillance is to ensure that effective programs have been developed and implemented to protect the public and DOE's workers from unplanned criticality. The programs should...

  5. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect (OSTI)

    Lawrence, J.D.

    1993-11-01

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

  6. 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.

  7. 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.

  8. 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.

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

    Office of Environmental Management (EM)

    & Technology 2 Outline Nuclear Safety Research & Development Overview Summary of EM- NSR&D Presentations from February 2009 Evaluating Performance of Nuclear Grade HEPA Filters ...

  10. 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...

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

    Energy Savers [EERE]

    Nuclear Safety: Safety Basis and Quality Assurance at the Los Alamos National Laboratory DOE/IG-0941 July 2015 U.S. Department of Energy Office of Inspector General Office of Audits and Inspections Department of Energy Washington, DC 20585 July 16, 2015 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman Inspector General SUBJECT: INFORMATION: Audit Report: "Follow-up on Nuclear Safety: Safety Basis and Quality Assurance at the Los Alamos National Laboratory" BACKGROUND A primary

  12. Plutonium Pits | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Plutonium Pits Plutonium pits are a critical core component of a nuclear weapon. To ensure the reliability, safety, and security of nuclear weapons without underground nuclear ...

  13. Plutonium Pits | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Pits Plutonium pits are a critical core component of a nuclear weapon. To ensure the reliability, safety, and security of nuclear weapons without underground nuclear testing;...

  14. DOE Spent Nuclear Fuel Group in Support of Criticality, DBE, TSPA-LA

    SciTech Connect (OSTI)

    Henry Loo

    2000-05-01

    This report presents the basis for grouping the over 250 Department of Energy (DOE) spent nuclear fuel (SNF) types in support of analyses for final repository disposal. For each of the required analyses, the parameters needed in conducting the analyses were identified and reviewed. The grouping proposed for the three types of analyses (criticality, design basis events, and total system performance assessment) are based on the similarities of DOE SNF as a function of these parameters. As necessary, further justifications are provided to further reduce the DOE SNF grouping in support of the Office of Civilian Radioactive Waste Management Systems preclosure and postclosure safety cases.

  15. 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.

  16. 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.

  17. Office of Safety Infrastructure & Operations | National Nuclear Security

    National Nuclear Security Administration (NNSA)

    Administration Safety Infrastructure & Operations NNSA's G2 Management Information System Wins Association for Enterprise Information's (AFEI) "Excellence in Enterprise Information Award" The G2 team and the 2015 Association for Enterprise Information's (AFEI) Excellence in Enterprise Information Award. (WASHINGTON, D.C) - The National Nuclear Security Administration (NNSA) has received the 2015 Association for Enterprise Information's (AFEI) Excellence in Enterprise

  18. 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.

  19. Safety

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

    - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced Nuclear Energy Nuclear

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

    Energy Savers [EERE]

    SENSITIVE DOE-STD-1104-2009 May 2009 Superseding DOE-STD-1104-96 DOE STANDARD REVIEW AND APPROVAL OF NUCLEAR FACILITY SAFETY BASIS AND SAFETY DESIGN BASIS DOCUMENTS U.S. Department of Energy AREA SAFT Washington, DC 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. DOE-STD-1104-2009 ii Available on the Department of Energy Technical Standards web page at http://www.hss.energy.gov/nuclearsafety/ns/techstds/ DOE-STD-1104-2009 iii CONTENTS FOREWORD

  1. 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.

  2. 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.

  3. 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.

  4. 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.

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

    Broader source: Energy.gov [DOE]

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

  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. 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.

  8. 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)

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

    Broader source: Energy.gov [DOE]

    Defense Nuclear Facilities Safety Board Public Meeting on the Status of Integration of Safety Into the Design of the Uranium Processing Facility [HIAR-Y-12-2012-10-02

  10. 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)

  11. AIM-98-3464 RECEIVED THE HISTORY OF NUCLEAR WEAPON SAFETY DEVICES

    Office of Scientific and Technical Information (OSTI)

    AIM-98-3464 RECEIVED THE HISTORY OF NUCLEAR WEAPON SAFETY DEVICES David W. Plummer and William H. Greenwood Sandia National Laboratories Juri 0 8 8% Surety Components and Instrumentation Center Albuquerque, NM 87185-0329 Abstract 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

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

    Broader source: Energy.gov [DOE]

    This SRP, Nuclear Safety Basis Program Review, consists of five volumes. It provides information to help strengthen the technical rigor of line management oversight and federal monitoring of DOE nuclear facilities. It provides a primer on the safety basis development and documentation process used by the DOE. It also provides a set of LOIs for the review of safety basis programs and documents of nuclear facilities at various stages of the facility life cycle.

  13. DOE Cites Fluor Fernald Inc. for Nuclear Safety Violations | Department of

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

    Energy Fluor Fernald Inc. for Nuclear Safety Violations DOE Cites Fluor Fernald Inc. for Nuclear Safety Violations August 25, 2005 - 2:43pm Addthis Washington, D.C. - The Department of Energy (DOE) today notified Fluor Fernald, Inc. (Fluor Fernald) that it will fine the company $33,000 for violations of the department's nuclear safety requirements. Fluor Fernald is the department's contractor responsible for environmental cleanup activities at the Fernald Closure Project. The Preliminary

  14. DOE-STD-1185-2004; Nuclear Explosive Safety Study Functional...

    Office of Environmental Management (EM)

    STD-1185-2004 August 2004 DOE STANDARD NUCLEAR EXPLOSIVE SAFETY STUDY FUNCTIONAL AREA ... gravity bombs, including: * Radars * Contact fuses * Timers * Power supplies b. ...

  15. Nuclear Explosives Safety Evaluation Process (DOE-STD-3015-2004...

    Energy Savers [EERE]

    DOE-STD-3015-2001 DOE STANDARD NUCLEAR EXPLOSIVE SAFETY EVALUATION PROCESS U.S. Department of ... production, manufacturing, andor power plant operations. * Demonstrated ...

  16. CNS names Guess Director of Nuclear Safety Oversight | Y-12 National...

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

    CNS names Guess Director of ... CNS names Guess Director of Nuclear Safety Oversight Posted: ... Guess' most recent position was as Power Ascension Test Director in the ...

  17. DOE P 420.1 Department of Energy Nuclear Safety Policy, Approved: 2-08-2011

    Broader source: Energy.gov [DOE]

    PURPOSE: To document the Department of Energy’s (DOE) nuclear safety policy.SCOPE: The provisions of this policy apply to all Departmental elements with responsibility for a nuclear facility,...

  18. Cover letter, 10/29/03, re Nuclear Safety Technical Position, Deliverable 4.2.1

    Broader source: Energy.gov [DOE]

    The enclosed Nuclear Safety Technical Position is Deliverable 4.2.1. under the Implementation Plan for Defense Nuclear Facilitises Board (DNFSB) Recommendation 2002-3, Requirements for Design...

  19. 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.

  20. Enforcement handbook: Enforcement of DOE nuclear safety requirements

    SciTech Connect (OSTI)

    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.

  1. 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.

  2. 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.

  3. Office Of Nuclear Energy

    Office of Environmental Management (EM)

    the Dependability Attributes of Software-Based Safety Critical Instrumentation and Control Systems in Nuclear Power Plants) (Carol Smidts) (The Ohio State University) (NEET ...

  4. New Research Center to Increase Safety and Power Output of U.S. Nuclear

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

    Reactors | Department of Energy Research Center to Increase Safety and Power Output of U.S. Nuclear Reactors New Research Center to Increase Safety and Power Output of U.S. Nuclear Reactors May 3, 2011 - 3:41pm Addthis Oak Ridge, Tenn. - Today the Department of Energy dedicated the Consortium for Advanced Simulation of Light Water Reactors (CASL), an advanced research facility that will accelerate the advancement of nuclear reactor technology. CASL researchers are using supercomputers to

  5. 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)

  6. Criticality Safety Lessons Learned in a Deactivation and Decommissioning Environment [A Guide for Facility and Project Managers

    SciTech Connect (OSTI)

    NIRIDER, L.T.

    2003-08-06

    This document was designed as a reference and a primer for facility and project managers responsible for Deactivation and Decommissioning (D&D) processes in facilities containing significant inventories of fissionable materials. The document contains lessons learned and guidance for the development and management of criticality safety programs. It also contains information gleaned from occurrence reports, assessment reports, facility operations and management, NDA program reviews, criticality safety experts, and criticality safety evaluations. This information is designed to assist in the planning process and operational activities. Sufficient details are provided to allow the reader to understand the events, the lessons learned, and how to apply the information to present or planned D&D processes. Information is also provided on general lessons learned including criticality safety evaluations and criticality safety program requirements during D&D activities. The document also explores recent and past criticality accidents in operating facilities, and it extracts lessons learned pertinent to D&D activities. A reference section is included to provide additional information. This document does not address D&D lessons learned that are not pertinent to criticality safety.

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

    Broader source: Energy.gov [DOE]

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

  8. 2015 Call for Proposals for the Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program

    Broader source: Energy.gov [DOE]

    2015 Call for Proposals for the Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program.

  9. Energy Department and Catholic University Improve Safety of Nuclear...

    Energy Savers [EERE]

    The center provides support to various nuclear facilities in the United States, Japan and the United Kingdom. Hanford is a nuclear and chemical waste processing facility under ...

  10. DOE Standard on Development and Use of Probabilistic Risk Assessment in DOE Nuclear Safety Applications (draft), December 2010

    Broader source: Energy.gov [DOE]

    There have been significant developments with regard to the risk assessment and risk informed decision making, as it applies to nuclear and other safety areas, since the Department of Energy (DOE) developed its approach to managing nuclear safety. The developments and associated technical insights may be of use to DOE in its efforts to continuously improve safety performance at its nuclear facilities.

  11. 2016 Call for Proposals for the Department of Energy (DOE) Nuclear Safety

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

    Research and Development (NSR&D) Program | Department of Energy Call for Proposals for the Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program 2016 Call for Proposals for the Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program January 20, 2016 - 9:10am Addthis 2016 Call for Proposals for the Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program These documents include the Proposal

  12. DEPARTMENT OF ENERGY CITES FLUOR B&W PORTSMOUTH, LLC FOR NUCLEAR SAFETY AND

    Energy Savers [EERE]

    FLUOR B&W PORTSMOUTH, LLC FOR NUCLEAR SAFETY AND RADIATION PROTECTION VIOLATIONS January 30, 2015 - 4:14pm Share on emailShare on facebook NEWS MEDIA CONTACT * 202 586 4940 * DOENews@hq.doe.gov Department of Energy Cites Fluor B&W Portsmouth, LLC for Nuclear Safety and Radiation Protection Violations WASHINGTON, D.C. - The U.S. Department of Energy (DOE) today issued a Preliminary Notice of Violation (PNOV) to Fluor B&W Portsmouth (FBP) for violations of the DOE's nuclear safety and

  13. Long term nuclear criticality potential in waste packages

    SciTech Connect (OSTI)

    Thomas, D.A.; Doering, T.W.

    1994-12-31

    Title 10 CFR 60.131.(b).(7) requires that the radioactive waste disposed of in the Mined Geologic Disposal System (MGDS) remain subcritical during the period of isolation. The period of waste isolation, approximately 10,000 years, represents a time period greater than any previously examined for criticality control of spent fuel. Change in the criticality potential over long time periods for the Multi-Purpose Canister (MPC) waste package conceptual design has been examined and methods of criticality control over this time have been investigated.

  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. Criticality Safety Envelope for Receipt, Handling, and Storage of Transuranic Waste

    SciTech Connect (OSTI)

    Vincent, A.M.

    1998-12-04

    Current criticality safety limits for Solid Waste Management Facility (SWMF) Transuranic (TRU) Waste Storage Pads are based on analysis of systems where mass is the only independent parameter and all other parameters are assumed at their most reactive values (Ref. 1). These limits result in administrative controls (i.e., limit stacking of containers, coordination of drums for culvert storage based on individual drum fissile inventories, and mass limits for accumulation of polyethylene boxes in culverts) which can only be met by redundant SWMF administrative controls. These analyses did not credit the nature of the waste generator process that would provide bounding limits on the other parameters (i.e. less than optimal moderation and configurations within packages (containers)). They also did not indicate the margin of safety associated with operating to these mass limits. However, by crediting the waste generator processes (and maintaining such process assumptions via controls in the criteria for waste acceptance) sufficient margin of safety can be demonstrated to justify continued SWMF TRU pad operation with fewer administrative controls than specified in the Double Contingency analysis (DCA) (Ref. 1).

  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. DOE-STD-1183-2004; Nuclear Safety Specialist Functional Area...

    Energy Savers [EERE]

    DOE-STD-1183-2004 April 2004 DOE STANDARD NUCLEAR SAFETY SPECIALIST FUNCTIONAL AREA ... Given a system diagram, identify the power sources andor loads and their status. d. ...

  18. The history of nuclear weapon safety devices (Conference) | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    The history of nuclear weapon safety devices Citation Details In-Document Search Title: The history of nuclear weapon safety devices × You are accessing a document from the Department of Energy's (DOE) SciTech Connect. This site is a product of DOE's Office of Scientific and Technical Information (OSTI) and is provided as a public service. Visit OSTI to utilize additional information resources in energy science and technology. A paper copy of this document is also available for sale to the

  19. Safety Aspects of Wet Storage of Spent Nuclear Fuel, OAS-L-13-11

    Energy Savers [EERE]

    Safety Aspects of Wet Storage of Spent Nuclear Fuel OAS-L-13-11 July 2013 Department of Energy Washington, DC 20585 July 10, 2013 MEMORANDUM FOR THE SENIOR ADVISOR FOR ENVIRONMENTAL MANAGEMENT FROM: Daniel M. Weeber Assistant Inspector General for Audits and Administration Office of Inspector General SUBJECT: INFORMATION: Audit Report on "Safety Aspects of Wet Storage of Spent Nuclear Fuel" BACKGROUND The Department of Energy (Department) is responsible for managing and storing spent

  20. National Nuclear Security Administration Fact Sheet Preliminary Notice of Violation: Worker Safety and Health Violations at

    Energy Savers [EERE]

    Nuclear Security Administration Fact Sheet Preliminary Notice of Violation: Worker Safety and Health Violations at Y-12 National Security Complex On May 27, 2015, the National Nuclear Security Administration (NNSA) issued a Preliminary Notice of Violation (PNOV) to Babcock & Wilcox Technical Services Y-12, LLC (B&W Y-12) for violations of Department of Energy (DOE) worker safety and health program requirements. At the time of the events, B&W Y-12 was the management and operating

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

    SciTech Connect (OSTI)

    1995-12-01

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

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

    SciTech Connect (OSTI)

    1995-12-01

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

  3. 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).

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

    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) 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.

  5. 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.

  6. Criticality safety evaluation for long term storage of FFTF fuel in interim storage casks

    SciTech Connect (OSTI)

    Richard, R.F.

    1995-05-11

    It has been postulated that a degradation phenomenon, referred to as ``hot cell rot``, may affect irradiated FFTF mixed plutonium-uranium oxide (MOX) fuel during dry interim storage. ``Hot cell rot`` refers to a variety of phenomena that degrade fuel pin cladding during exposure to air and inert gas environments. It is thought to be a form of caustic stress corrosion cracking or environmentally assisted cracking. Here, a criticality safety analysis was performed to address the effect of the ``hot cell rot`` phenomenon on the long term storage of irradiated FFTF fuel in core component containers. The results show that seven FFTF fuel assemblies or six Ident-69 pin containers stored in core component containers within interim storage casks will remain safely subcritical.

  7. 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

  8. 2015 Nuclear and Facility Safety Programs Workshop Block Agenda

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

    Worthington (AU-10) ISM o Colette Broussard (AU-23) QA and Other Data Trending o Pat Lewis (SC-CH) Lessons From Safety Basis Reviews o Carl Sykes (NA-511) Exemptions * Readiness...

  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. DOE's Nuclear Weapons Complex: Challenges to Safety, Security, and Taxpayer Stewardship

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

    Oversight and Investigations 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 10:00 AM September 12, 2012 1 Mr. Chairman and Members of the Subcommittee, I am pleased to be here at your request to testify on matters relating to the Department of Energy's oversight of the nuclear weapons complex. 1 The National Nuclear Security Administration (NNSA) was

  11. Nuclear Systems Design and Analysis (NSD&A) Division | Department of Energy

    Office of Environmental Management (EM)

    Nuclear Safety Enforcement Documents Nuclear Safety Enforcement Documents Documents Available for Download April 28, 2016 Savannah River Nuclear Solutions, LLC, Consent Order NCO-2016-01 Nuclear Safety Enforcement Consent Order issued to Savannah River Nuclear Solutions, LLC relating to nuclear criticality safety infractions that occurred at the Savannah River Site February 19, 2016 Preliminary Notice of Violation, Los Alamos National Security, LLC Nuclear Safety Enforcement Preliminary Notice

  12. 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.

  13. An advanced deterministic method for spent-fuel criticality safety analysis

    SciTech Connect (OSTI)

    DeHart, M.D.

    1998-09-01

    Over the past two decades, criticality safety analysts have come to rely to a large extent on Monte Carlo methods for criticality calculations. Monte Carlo has become popular because of its capability to model complex, nonorthogonal configurations or fissile materials, typical of real-world problems. In the last few years, however, interest in determinist transport methods has been revived, due to shortcomings in the stochastic nature of Monte Carlo approaches for certain types of analyses. Specifically, deterministic methods are superior to stochastic methods for calculations requiring accurate neutron density distributions or differential fluxes. Although Monte Carlo methods are well suited for eigenvalue calculations, they lack the localized detail necessary to assess uncertainties and sensitivities important in determining a range of applicability. Monte Carlo methods are also inefficient as a transport solution for multiple-pin depletion methods. Discrete ordinates methods have long been recognized as one of the most rigorous and accurate approximations used to solve the transport equation. However, until recently, geometric constrains in finite differencing schemes have made discrete ordinates methods impractical for nonorthogonal configurations such as reactor fuel assemblies. The development of an extended step characteristic (ESC) technique removes the grid structure limitation of traditional discrete ordinates methods. The NEWT computer code, a discrete ordinates code built on the ESC formalism, is being developed as part of the SCALE code system. This paper demonstrates the power, versatility, and applicability of NEWT as a state-of-the-art solution for current computational needs.

  14. 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.

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

    Energy Savers [EERE]

    ... nuclear facilities-PF-4, the Weapons Engineering Tritium Facility (WETF), and the Radioassay and Nondestructive Testing ... review identified a long history of operational issues that ...

  16. 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.

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

    SciTech Connect (OSTI)

    Pinkston, David

    2015-02-24

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

  18. Nuclear Reaction Cross Sections Database at BNL | U.S. DOE Office...

    Office of Science (SC) Website

    advanced reactor and fuel cycle concepts, criticality safety, fusion, space applications, nuclear astrophysics and nuclear physic facility design. The ENDFB-VII.0 ...

  19. 2016 Strategic Plan Chief of Nuclear Safety | Department of Energy

    Office of Environmental Management (EM)

    National Fall Prevention Campaign 2016 National Fall Prevention Campaign March 17, 2016 - 9:07am Addthis 2016 National Fall Prevention Campaign As part of a Fall Prevention event initiated by the Occupational Safety and Health Administration (OSHA), the 3rd Annual National Fall Prevention Campaign will take place on May 2-6. This event is a nationwide voluntary effort to remind and educate employers and workers in the construction industry of the serious dangers regarding falls from elevated

  20. National Nuclear Security Administration Fact Sheet Preliminary Notice of Violation: Worker Safety and Health Violations at

    Energy Savers [EERE]

    Sandia National Laboratories On November 17, 2015, the National Nuclear Security Administration (NNSA) issued a Preliminary Notice of Violation (PNOV) to Sandia Corporation (Sandia) for violations of Department of Energy (DOE) worker safety and health requirements. Sandia is the management and operating contractor for NNSA's Sandia National Laboratories (SNL) located in Albuquerque, New Mexico. The PNOV cites Sandia for deficiencies in implementation of worker safety and health program

  1. Intact and Degraded Criticality Calculations for the Codisposal of Shippingport LWBR Spent Nuclear Fuel in a Waste Package

    SciTech Connect (OSTI)

    L.M. Montierth

    2000-09-15

    The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the U.S. Department of Energy's (DOE) Shippingport Light Water Breeder Reactor (SP LWBR) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP), which is to be placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (K{sub eff}) for intact- and degraded-mode internal configurations of the codisposal WP containing Shippingport LWBR seed-type assemblies. The results of this calculation will be used to evaluate criticality issues and support the analysis that is planed to be performed to demonstrate the viability of the codisposal concept for the MGR. This calculation is associated with the waste package design and was performed in accordance with the DOE SNF Analysis Plan for FY 2000 (See Ref. 22). The document has been prepared in accordance with the Administrative Procedure AP-3.12Q, Calculations (Ref. 23).

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

    SciTech Connect (OSTI)

    George, T.G.

    1989-02-01

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

  3. 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.

  4. Criticality safety review of 2 1/2 -, 10-, and 14-ton UF sub 6 cylinders

    SciTech Connect (OSTI)

    Broadhead, B.L.

    1991-01-01

    The US regulations governing the packaging and transportation of UF{sub 6} cylinders are contained in the publication 10CFR71. Under the current 10CFR71 regulations, packages are classified according to Fissile Class I, II, or III and a corresponding transport index (TI). UF{sub 6} cylinders designed to contain 2{1/2}-tons of UF{sub 6} are classified as Fissile Class II packages with a TI of 5 for the purpose of transportation. The 10-ton UF{sub 6} cylinders are classified as Fissile Class I with no TI assigned for transportation. The 14-ton cylinders are not certified for transport with enrichments greater than 1 wt % since they have no approved overpack. This work reviews the suitability of 2{1/2}-ton UF{sub 6} packages for reclassification as Fissile Class I with a maximum {sup 235}U enrichment of 5 wt %. Additionally, the 10- and 14-ton cylinders are reviewed to address a change in maximum {sup 235}U enrichment from 4.5 to 5 wt %. Based on this evaluation, the 2{1/2}-ton UF{sub 6} cylinders meet the 10CFR71 criteria for Fissile Class I packages, and no TI is needed for criticality safety purposes. Similarly, the 10- and 14-ton UF{sub 6} packages appear suitable for a maximum enrichment rating change to 5 wt % {sup 235}U. 6 refs., 4 figs., 1 tab.

  5. Nuclear Safety Information Agreement Between the U.S. Nuclear Regulatory

    Energy Savers [EERE]

    of Energy Regulatory Commission Regulatory and Licensing Matters Nuclear Regulatory Commission Regulatory and Licensing Matters GC-52 provides legal advice to DOE regarding Nuclear Regulatory Commission (NRC) regulatory and licensing matters of interest to DOE, either as an NRC license applicant or in connection with related authorities and responsibilities of DOE and NRC on nuclear material, nuclear waste, and nuclear nonproliferation matters. GC-52 attorneys provide advice and support on a

  6. 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.

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

    DOE Patents [OSTI]

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

    1991-01-01

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

  8. 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.

  9. 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.

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

    SciTech Connect (OSTI)

    D. Kokkinos

    2005-04-28

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

  11. 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 costs of code data library updating, the use of S/U analysis methodologies could be accomplished on a shorter schedule and a lower cost than the gathering of sufficient experimental data. ENERCON estimates of the costs of an updated S/U computer code and data suite are $5M to $10M with a schedule of two to three years. Recent ORNL analyses using the S/U analysis method show that the bias and uncertainty values for fission product cross sections are smaller than previously expected. This result is confirmed by a similar EPRI approach using different data and computer codes. ENERCON also found that some issues regarding the implementation of burnup credit appear to have been successfully resolved especially the axial burnup profile issue and the depletion parameter issue. These issues were resolved through data gathering activities at the Yucca Mountain Project and ORNL.

  12. Safety

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

    Safety All JLF participants must comply fully with all LLNL safety regulations and procedures by becoming a Registered User of the facility. All JLF participants must complete available LLNL safety training: HS5200-W Laser Safety HS4258-W Beryllium Awareness HS4261-W Lead Awareness HS5220-W Electrical Safety Awareness HS6001-W 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

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

    SciTech Connect (OSTI)

    Cottrell, W.B.

    2001-08-17

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

  14. 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.

  15. 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.

  16. 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 Departments 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 Departments 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.

  17. 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)

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

    SciTech Connect (OSTI)

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

    1983-12-01

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

  19. National Nuclear Security Administration Fact Sheet Preliminary Notice of Violation: Worker Safety and Health Violations at

    Energy Savers [EERE]

    Nevada National Security Site On August 25, 2015, the National Nuclear Security Administration (NNSA) issued a Preliminary Notice of Violation (PNOV) to National Security Technologies, LLC (NSTec) for violations of Department of Energy (DOE) worker safety and health requirements. NSTec is the management and operating contractor to DOE's NNSA Nevada National Security Site (NNSS). The violations are associated with the chemical explosion that injured two workers at NNSS's Nonproliferation Test and

  20. NNSA lab makes fire tornados to ensure weapon safety | National Nuclear

    National Nuclear Security Administration (NNSA)

    Security Administration lab makes fire tornados to ensure weapon safety Tuesday, May 17, 2016 - 11:33am Fire whirls from a 3-meter diameter pool in the Thermal Test Complex at Sandia National Laboratories. NNSA's laboratories need to conduct a lot of tests under extreme conditions. From enormous implosion pressure to energy more intense than the sun, NNSA's labs support national security by verifying the U.S. nuclear weapons stockpile without underground explosive testing. In Sandia National

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

    SciTech Connect (OSTI)

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

    1991-01-01

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

  2. [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...

  3. Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports

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

    1997-12-12

    he purpose of this DOE Standard is to establish guidance for the preparation and review of hazard categorization and accident analyses techniques as required in DOE Order 5480.23, Nuclear Safety Analysis Reports.

  4. Microsoft PowerPoint - Fire Safety workshop NQA-1 CGD 4 29 15 [Read-Only]

    Office of Environmental Management (EM)

    v Nuclear and Facility Safety Programs Workshop Fire Safety Track May 5th, 2015 Overview NQA 1 Commercial Grade Dedication Critical Characteristics Department of Energy Nuclear and Facility Safety Programs Workshop Fire Safety Track May 5 th , 2015 Randy P. Lanham PE, CSP Dale Moon, PE Fire Protection Chief Engineer Facility Engineering Depart. Mng. Randy.Lanham@cns.doe.gov Consolidated Nuclear Security Pantex and Y12 2 Overview CGD Definition Safety Function / DSA Requirements Example of CGD

  5. Management of the aging of critical safety-related concrete structures in light-water reactor plants

    SciTech Connect (OSTI)

    Naus, D.J.; Oland, C.B. ); Arndt, E.G. )

    1990-01-01

    The Structural Aging Program has the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant safety-related structures for continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued-service determinations. Objectives, accomplishments, and planned activities under each of these tasks are presented. Major program accomplishments include development of a materials property data base for structural materials as well as an aging assessment methodology for concrete structures in nuclear power plants. Furthermore, a review and assessment of inservice inspection techniques for concrete materials and structures has been complete, and work on development of a methodology which can be used for performing current as well as reliability-based future condition assessment of concrete structures is well under way. 43 refs., 3 tabs.

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

    Broader source: Energy.gov [DOE]

    In the six months after the March 2011 Fukushima Daiichi nuclear power plant accident in Japan, the U.S. Department of Energy (DOE) took several actions to review the safety of its nuclear facilities and identify situations where near-term improvements could be made.

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

    SciTech Connect (OSTI)

    Gilli, Ludivine; Charron, Sylvie

    2012-07-01

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

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

    SciTech Connect (OSTI)

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

    2015-07-15

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

  9. 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.

  10. 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.

  11. Enterprise Assessments Lessons Learned from Targeted Reviews of the Management of Safety Systems at U.S. Department of Energy Nuclear Facilities … April 2016

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

    Office of Enterprise Assessments Lessons Learned from Targeted Reviews of the Management of Safety Systems at U.S. Department of Energy Nuclear Facilities April 2016 Office of Nuclear Safety and Environmental Assessments Office of Environment, Safety and Health Assessments Office of Enterprise Assessments U.S. Department of Energy i Table of Contents Acronyms

  12. 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.

  13. 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.

  14. 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.

  15. Health and safety impacts related to the management of spent nuclear fuels

    SciTech Connect (OSTI)

    Jilek, D.C.

    1996-06-01

    Under the Nuclear Waste Policy Act of 1982, as amended, the U.S. Department of Energy is responsible for managing the disposal of spent nuclear fuel from civilian nuclear power plants. Deployment of a multipurpose canister (MPC) system for dry storage of commercial spent nuclear fuel at reactor sites was determined to be an option for managing spent nuclear fuel until either a permanent repository or interim central storage facility (commonly called a Monitored Retrievable Storage Facility, or MRS) becomes available. Routine health and safety impacts to workers from handling and storage operations at nuclear facilities for four separate scenarios were evaluated for the MPC system: an on-time repository with an MRS; an on-time repository with no MRS; a delayed repository with an MRS; and a delayed repository with no MRS. In addition to evaluating the MPC system, five alternatives were analyzed. These included the No Action Alternative (NAA), Current Technology (CTr), the Transposable Storage Cask (TSC), the Dual-Purpose Canister (DPC), and the Small MPC (SmMPC). Health effects are expressed as collective doses in person- rem per year and risks as latent cancer fatalities per year for incident-free operations for each alternative and scenario. Results show that both dose and risks to workers vary as much as 68{percent} among scenarios and alternatives. Although dose estimates and risks fall below limits for radiation dose to workers as specified in Title 10, Part 20, of the Code of Federal Regulations, additional measures could be applied to reduce potential doses and resultant health risk. 5 refs., 2 tabs.

  16. 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.

  17. Research on long term safety of nuclear waste disposal at the research center Karlsruhe, Germany

    SciTech Connect (OSTI)

    Gompper, Klaus; Bosbach, Dirk; Denecke, Melissa A.; Geckeis, Horst; Kienzler, Bernhard; Klenze, Reinhardt

    2007-07-01

    In Germany the safe disposal of radioactive waste is in the responsibility of the federal government. The R and D performed in the Institute for Nuclear Waste Disposal (INE) at the Research Center Karlsruhe contributes to the German provident research in the field of long-term safety for final disposal of high level heat producing nuclear wastes. INE's research is focused on the actinide elements and long lived fission products since these dominate the radiotoxicity over a long time. The research strategy synergistically combines fundamental science of aquatic radionuclide chemistry with applied investigations of real systems (waste form, host rock, aquifer), studied on laboratory scale and in underground laboratories. Because Germany has not yet selected a site for a high-level waste repository, all host rock formations under discussion in the international community (salt, hard rock, clay/tone) are investigated. Emphasis in long-term safety R and D at INE is on the development of actinide speciation methods and techniques in the trace concentration range. (authors)

  18. 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.

  19. 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)

  20. 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.

  1. 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.

  2. 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.

  3. 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.

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

    SciTech Connect (OSTI)

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

    1987-11-24

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

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

    Broader source: Energy.gov [DOE]

    In the six months after the March 2011 Fukushima Daiichi nuclear power plant accident in Japan, the U.S. Department of Energy (DOE) took several actions to review the safety of its nuclear facilities and identify situations where near-term improvements could be made. These actions and recommendations were addressed in an August 2011 report to the Secretary of Energy, Review of Requirements and Capabilities for Analyzing and Responding to Beyond Design Basis Events.

  6. 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...

  7. 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.

  8. Our Leadership | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Leadership The NNSA plays a critical role in ensuring the security of our Nation by maintaining the safety, security, and effectiveness of the U.S. nuclear weapons stockpile without nuclear testing; reducing the global danger from the proliferation of nuclear weapons and materials; providing the U.S. Navy with safe and effective nuclear propulsion; and providing the Nation with an effective nuclear counterterrorism and incident response capability. The NNSA plays a critical role in ensuring the

  9. Validation Study for Crediting Chlorine in Criticality Analyses for US Spent Nuclear Fuel Disposition

    SciTech Connect (OSTI)

    Sobes, Vladimir; Scaglione, John M.; Wagner, John C.; Dunn, Michael E.

    2015-01-01

    Spent nuclear fuel (SNF) management practices in the United States rely on dry storage systems that include both canister- and cask-based systems. The United States Department of Energy Used Fuel Disposition Campaign is examining the feasibility of direct disposal of dual-purpose (storage and transportation) canisters (DPCs) in a geological repository. One of the major technical challenges for direct disposal is the ability to demonstrate the subcriticality of the DPCs loaded with SNF for the repository performance period (e.g., 10,000 years or more) as the DPCs may undergo degradation over time. Specifically, groundwater ingress into the DPC (i.e., flooding) could allow the system to achieve criticality in scenarios where the neutron absorber plates in the DPC basket have degraded. However, as was shown by Banerjee et al., some aqueous species in the groundwater provide noticeable reactivity reduction for these systems. For certain amounts of particular aqueous species (e.g., chlorine, lithium) in the groundwater, subcriticality can be demonstrated even for DPCs with complete degradation of the neutron absorber plates or a degraded fuel basket configuration. It has been demonstrated that chlorine is the leading impurity, as indicated by significant neutron absorption in the water that is available in reasonable quantities for the deep geological repository media under consideration. This paper presents the results of an investigation of the available integral experiments worldwide that could be used to validate DPC disposal criticality evaluations, including credit for chlorine. Due to the small number of applicable critical configurations, validation through traditional trending analysis was not possible. The bias in the eigenvalue of the application systems due only to the chlorine was calculated using TSURFER analysis and found to be on the order of 100 percent mille (1 pcm = 10-5 keff). This study investigated the design of a series of critical configurations with varying amounts of chlorine to address validation gaps. Such integral experiments would support the crediting of the chlorine neutron-absorption properties in groundwater and the demonstration of subcriticality for DPCs in deep geologic repositories with sufficient chlorine availability.

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

    DOE Patents [OSTI]

    Gabor, J.D.; Cassulo, J.C.; Pedersen, D.R.; Baker, L. Jr.

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and can be discharged from the reactor core. The invention provides a porous bed of sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

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

    DOE Patents [OSTI]

    Gabor, John D.; Cassulo, John C.; Pedersen, Dean R.; Baker, Jr., Louis

    1986-01-01

    The invention teaches safety apparatus that can be included in a nuclear reactor, either when newly fabricated or as a retrofit add-on, that will minimize proliferation of structural damage to the reactor in the event the reactor is experiencing an overheating malfunction whereby radioactive nuclear debris might break away from and be discharged from the reactor core. The invention provides a porous bed or sublayer on the lower surface of the reactor containment vessel so that the debris falls on and piles up on the bed. Vapor release elements upstand from the bed in some laterally spaced array. Thus should the high heat flux of the debris interior vaporize the coolant at that location, the vaporized coolant can be vented downwardly to and laterally through the bed to the vapor release elements and in turn via the release elements upwardly through the debris. This minimizes the pressure buildup in the debris and allows for continuing infiltration of the liquid coolant into the debris interior.

  12. Department of Energy Cites Fluor B&W Portsmouth, LLC for Nuclear Safety and Radiation Protection Violations

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) today issued a Preliminary Notice of Violation (PNOV) to Fluor B&W Portsmouth (FBP) for violations of the DOE’s nuclear safety and radiation protection regulations, and has proposed a $243,750 civil penalty.

  13. Safety, Security

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

    Safety, Security Safety, Security The Lab's mission is to develop and apply science and technology to ensure the safety, security, and reliability of the U.S. nuclear deterrent; reduce global threats; and solve other emerging national security and energy challenges. Contact Operator Los Alamos National Laboratory (505) 667-5061 We do not compromise safety for personal, programmatic, or operational reasons. Safety: we integrate safety, security, and environmental concerns into every step of our

  14. A probabilistic evaluation of the safety of Babcock and Wilcox nuclear reactor power plants with emphasis on historically observed operational events

    SciTech Connect (OSTI)

    Hsu, C.J.; Youngblood, R.W.; Fitzpatrick, R.G.; Amico, P.J.

    1989-03-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Reactor Regulation, Division of Engineering and System Technology (A/D for Systems), US Nuclear Regulatory Commission. This study was requested by the NRC to assist their staff in assessing the risk significance of features of the Babcock and Wilcox (B and W) reactor plant design in the light of recent operational events. This study focuses on a critical review of submissions from the B and W Owners Group (BWOG) and as an independent assessment of the risk significance of ''Category C'' events at each operating B and W reactor. Category C events are those in which system conditions reach limits which require significant safety system and timely operator response to mitigate. A precursor study for each of the major B and W historical Category C events also was carried out. In addition, selected PRAs for B and W reactor plants and plants with other pressurized water reactor (PWR) designs were reviewed to appraise their handling of Category C events, thereby establishing a comparison between the risk profiles of B and W reactor plants and those of other PWR designs. The effectiveness of BWOG recommendations set forth in Appendix J of the BWOG SPIP (Safety and Performance Improvement Program) report (BAW-1919) also was evaluated. 49 refs., 21 figs., 52 tabs.

  15. Criticality Safety Review of 2 1/2-, 10-, and 14-Ton UF(Sub 6) Cylinders

    SciTech Connect (OSTI)

    Broadhead, B.L.

    1991-01-01

    Currently, UF{sub 6} cylinders designed to contain 2 1/2 tons of UF{sub 6} are classified as Fissile Class II packages with a transport index (TI) of 5 for the purpose of transportation. The 10-ton UF{sub 6} cylinders are classified as Fissile Class I with no TI assigned for transportation. The 14-ton cylinders, although not certified for transport with enrichments greater than 1 wt % because they have no approved overpack, can be used in on-site operations for enrichments greater than 1 wt %. The maximum 235U enrichments for these cylinders are 5.0 wt % for the 2 1/2-ton cylinder and 4.5 wt % for the 10- and 14-ton cylinders. This work reviews the suitability for reclassification of the 2 1/2-ton UF{sub 6} packages as Fissile Class I with a maximum {sup 235}U enrichment of 5 wt %. Additionally, the 10- and 14-ton cylinders are reviewed to address a change in maximum {sup 235}U enrichment from 4.5 to 5 wt %. Based on this evaluation, the 2 1/2-ton UF{sub 6} cylinders meet the 10 CFR.71 criteria for Fissile Class I packages, and no TI is needed for criticality safety purposes; however, a TI may be required based on radiation from the packages. Similarly, the 10- and 14-ton UF{sub 6} packages appear acceptable for a maximum enrichment rating change to 5 wt % {sup 235}U.

  16. 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.

  17. 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

  18. 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...

  19. 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

  20. 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

  1. 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.

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

    SciTech Connect (OSTI)

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

    2013-10-01

    A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis and 2) topology-based methodologies to interactively visualize multidimensional data and extract risk-informed insights. Regarding item 1) we employ learning algorithms that aim to infer/predict simulation outcome and decide the coordinate in the input space of the next sample that maximize the amount of information that can be gained from it. Such methodologies can be used to both explore and exploit the input space. The later one is especially used for safety analysis scopes to focus samples along the limit surface, i.e. the boundaries in the input space between system failure and system success. Regarding item 2) we present a software tool that is designed to analyze multi-dimensional data. 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.

  3. 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.

  4. Canister storage building evaluation of nuclear safety for solidified high-level waste transfer and storage

    SciTech Connect (OSTI)

    Kidder, R.J., Westinghouse Hanford

    1996-09-17

    This document is issued to evaluate the safety impacts to the Canister Storage Building from transfer and storage of solidified high-level waste.

  5. 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)

  6. Strategic Safety Goals | Department of Energy

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

    Safety Goals More Documents & Publications Strategic Safety Goals Occupational Safety Performance Trends Development of the Nuclear Safety Information Dashboard - September 2012

  7. Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates

    SciTech Connect (OSTI)

    Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

    1996-05-01

    This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

  8. Transportation Safety

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

    Safety - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced Nuclear Energy

  9. Facility Safety | Department of Energy

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

    Facility Safety Facility Safety In addition to establishing nuclear safety requirements related to safety management programs that are essential to the safety of DOE nuclear facilities, the U.S. Department of Energy's (DOE) Office of Nuclear Facility Safety works proactively with headquarters and field offices to foster continuous improvement and nuclear safety excellence. In addition, the Office provides high quality, customer-oriented assistance that enables improved DOE program and field

  10. Facility Safety | Department of Energy

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

    Facility Safety Facility Safety In addition to establishing nuclear safety requirements related to safety management programs that are essential to the safety of DOE nuclear facilities, the U.S. Department of Energy's (DOE) Office of Nuclear Facility Safety works proactively with headquarters and field offices to foster continuous improvement and nuclear safety excellence. In addition, the Office provides high quality, customer-oriented assistance that enables improved DOE program and field

  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 research programs sponsored by Office of Nuclear Regulatory Research. Quarterly progress report, January 1-March 31, 1983. Volume 3, No. 1

    SciTech Connect (OSTI)

    Bari, R A; Cerbone, R J; Ginsberg, T; Greene, G A; Guppy, J G; Hall, R E; Luckas, Jr, W J; Reich, M; Saha, P; Sastre, C

    1983-06-01

    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; 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, Soil Structure Interaction; Human Error Data for Nuclear Power Plant Safety Related Events, Criteria for Human Engineering Regulatory Guides and Human Factors in Nuclear Power Plant Safeguards.

  13. Guidelines for nuclear power plant safety issue prioritization information development. Supplement 4

    SciTech Connect (OSTI)

    Tabatabai, A.S.; Fecht, B.A.; Powers, T.B.; Bickford, W.E.; Andrews, W.B.; Gallucci, R.H.V.; Bian, S.H.; Daling, P.M.; Eschbach, E.J.; Allen, C.H.

    1986-07-01

    This is the fifth in a series of reports to document the use of a methodology developed by the Pacific Northwest Laboratory to calculate, for prioritization purposes, the risk, dose and cost impacts of implementing resolutions to reactor safety issues (NUREG/CR-2800, Andrews et al. 1983). This report contains results of issue-specific analyses for 23 issues. Each issue was considered within the constraints of available information as of winter 1986, and two staff-weeks of labor. The results are referenced, as one consideration in setting priorities for reactor safety issues, in NUREG-0933, ''A Prioritization of Generic Safety Issues.''

  14. 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.

  15. 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.

  16. U.S., China Advance Nuclear Safety and Security Cooperation through...

    National Nuclear Security Administration (NNSA)

    General Hao. "Under the PUNT framework, both sides continue to promote effective and efficient measures to enhance peaceful uses of nuclear energy and strengthen public acceptance ...

  17. DOE-STD-0100T; DOE Standard Licensed Reactor Nuclear Safety Criteria...

    Energy Savers [EERE]

    ... Systems and Safety Criteria 7.2 Reactor Trip System ... RADIOACTIVE WASTE MANAGEMENT 11.1 Source Terms 11.2 Liquid Waste ... Exposures Are As Low As Reasonably Achievable ...

  18. 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.

  19. 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. 19

    SciTech Connect (OSTI)

    1995-11-01

    Supplement No. 19 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 (1) additional information submitted by the applicant since Supplement No. 18 was issued, and (2) matters that the staff had under review when Supplement No. 18 was issued.

  20. Program Analyst (Transportation Safety)

    Broader source: Energy.gov [DOE]

    A successful candidate in this position will serve as a Program Analyst(Transportation Safety) supporting and advising management on safety and health matters for nuclear and non-nuclear activities.

  1. 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.

  2. 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.

  3. 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.

  4. Review guidelines on software languages for use in nuclear power plant safety systems. Final report

    SciTech Connect (OSTI)

    Hecht, H.; Hecht, M.; Graff, S.; Green, W.; Lin, D.; Koch, S.; Tai, A.; Wendelboe, D.

    1996-06-01

    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: Ada, C/C++, Programmable Logic Controller (PLC) Ladder Logic, International Electrotechnical Commission (IEC) Standard 1131-3 Sequential Function Charts, Pascal, and PL/M. Appendices to the report include a tabular summary of the guidelines and additional information on selected languages.s

  5. 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).

  6. 2015 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.

  7. 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.

  8. 2012 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.

  9. 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.

  10. 2010 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

  11. 2011 Annual Workforce Analysis and Staffing Plan Report- Chief of Nuclear Safety

    Office of Energy Efficiency and Renewable Energy (EERE)

    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.

  12. Site Information | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    engineering, safety and reliability of the nuclear explosives package in nuclear weapons. ... engineering, safety and reliability of the nuclear explosives package in nuclear weapons. ...

  13. 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.

  14. 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)

  15. 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.

  16. Criticality concerns in cleaning large uranium hexafluoride cylinders

    SciTech Connect (OSTI)

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

    1995-06-01

    Cleaning large cylinders used to transport low-enriched uranium hexafluoride (UF{sub 6}) presents several challenges to nuclear criticality safety. This paper presents a brief overview of the cleaning process, the criticality controls typically employed and their bases. Potential shortfalls in implementing these controls are highlighted, and a simple example to illustrate the difficulties in complying with the Double Contingency Principle is discussed. Finally, a summary of recommended criticality controls for large cylinder cleaning operations is presented.

  17. Sandia Energy Nuclear Energy

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

    afety-expert-elected-to-national-academy-of-engineeringfeed 0 Sandia Teaches Nuclear Safety Course http:energy.sandia.govsandia-teaches-nuclear-safety-course http:...

  18. 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.

  19. 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

    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.

  20. 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.

  1. 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.

  2. Assessment of inservice conditions of safety-related nuclear plant structures

    SciTech Connect (OSTI)

    Ashar, H.; Bagchi, G.

    1995-06-01

    The report is a compilation from a number of sources of information related to the condition Of structures and civil engineering features at operating nuclear power plants in the United States. The most significant information came from the hands-on inspection of the six old plants (licensed prior to 1977) performed by the staff of the Civil Engineering and Geosciences Branch (ECGB) in the Division of Engineering of the Office of Nuclear Reactor Regulation. For the containment structures, most of the information related to the degraded conditions came from the licensees as part of the Licensing Event Report System (10 CFR 50.73), or as part of the requirement under limiting condition of operation of the plant-specific Technical Specifications. Most of the information related to the degradation of other Structures and civil engineering features was extracted from the industry survey, the reported incidents, and the plant visits. The report discusses the condition of the structures and civil engineering features at operating nuclear power plants and provides information that would help detect, alleviate, and correct the degraded conditions of the structures and civil engineering features.

  3. 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.

  4. Challenges to Integration of Safety and Reliability with Proliferation Resistance and Physical Protection for Generation IV Nuclear Energy Systems

    SciTech Connect (OSTI)

    H. Khalil; P. F. Peterson; R. Bari; G. -L. Fiorini; T. Leahy; R. Versluis

    2012-07-01

    The optimization of a nuclear energy system's performance requires an integrated consideration of multiple design goals - sustainability, safety and reliability (S&R), proliferation resistance and physical protection (PR&PP), and economics - as well as careful evaluation of trade-offs for different system design and operating parameters. Design approaches motivated by each of the goal areas (in isolation from the other goal areas) may be mutually compatible or in conflict. However, no systematic methodology approach has yet been developed to identify and maximize synergies and optimally balance conflicts across the possible design configurations and operating modes of a nuclear energy system. Because most Generation IV systems are at an early stage of development, design, and assessment, designers and analysts are only beginning to identify synergies and conflicts between PR&PP, S&R, and economics goals. The close coupling between PR&PP and S&R goals has motivated early attention within the Generation IV International Forum to their integrated consideration to facilitate the optimization of their effects and the minimization of potential conflicts. This paper discusses the status of this work.

  5. 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.

  6. Facility Safety - DOE Directives, Delegations, and Requirements

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

    0.1C Chg 1, Facility Safety by Pranab Guha Functional areas: DNFSB, Defense Nuclear Facility Safety and Health Requirement, Facility Safety, Requires Crosswalk When Revised,...

  7. 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.

  8. 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)

  9. 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.

  10. A safety and regulatory assessment of generic BWR and PWR permanently shutdown nuclear power plants

    SciTech Connect (OSTI)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A.

    1997-08-01

    The long-term availability of less expensive power and the increasing plant modification and maintenance costs have caused some utilities to re-examine the economics of nuclear power. As a result, several utilities have opted to permanently shutdown their plants. Each licensee of these permanently shutdown (PSD) plants has submitted plant-specific exemption requests for those regulations that they believe are no longer applicable to their facility. This report presents a regulatory assessment for generic BWR and PWR plants that have permanently ceased operation in support of NRC rulemaking activities in this area. After the reactor vessel is defueled, the traditional accident sequences that dominate the operating plant risk are no longer applicable. The remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times, could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. This study has defined four spent fuel configurations which encompass all of the anticipated spent fuel characteristics and storage modes following permanent shutdown. A representative accident sequence was chosen for each configuration. Consequence analyses were performed using these sequences to estimate onsite and boundary doses, population doses and economic costs. A list of candidate regulations was identified from a screening of 10 CFR Parts 0 to 199. The continued applicability of each regulation was assessed within the context of each spent fuel storage configuration and the results of the consequence analyses.

  11. Status and Effectiveness of DOE Efforts to Learn from Internal and External Operating Experience in Accordance with Commitment #20 of the DOE Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 2004-1

    Energy Savers [EERE]

    Safety and Security Report to the Secretary on the Status and Effectiveness of DOE Efforts to Learn from Internal and External Operating Experience in Accordance with Commitment #20 of the DOE Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 2004-1 February 2011 Office of Health, Safety and Security U.S. Department of Energy Office of Health, Safety and Security HSS Table of Contents 1.0 Introduction

  12. Test Site Operations & Maintenance Safety

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

    Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management ...

  13. 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.

  14. 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.

  15. National Security & Safety Reports | Department of Energy

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

    National Security & Safety Reports National Security & Safety Reports April 4, 2016 Audit Report: DOE-OIG-16-11 Followup on Western Area Power Administration's Critical Asset Protection December 15, 2015 Audit Report: OAI-L-16-05 Lithium Operations at the Y-12 National Security Complex September 1, 2015 Inspection Report: INS-L-15-03 Allegations Regarding Management of Highly Enriched Uranium July 16, 2015 Audit Report: DOE/IG-0941 Follow-up on Nuclear Safety: Safety Basis and Quality

  16. 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.

  17. safety analysis report

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

    Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering ...

  18. safety of space

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

    Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering ...

  19. Health, Safety, and Environmental Protection Committee Page 1

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

    Health, Safety, & Quality Assurance Health, Safety, & Quality Assurance Nuclear Safety and Worker Safety and Health training Nuclear Safety and Worker Safety and Health training PPPO's Safety and Health, Nuclear Safety, and Quality Assurance programs collectively ensure protection of public and worker health and safety and the environment. This is accomplished by empowering and holding accountable managers, employees and contractors to prioritize health, safety and environmental

  20. Health, Safety, & Quality Assurance | Department of Energy

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

    Health, Safety, & Quality Assurance Health, Safety, & Quality Assurance Nuclear Safety and Worker Safety and Health training Nuclear Safety and Worker Safety and Health training PPPO's Safety and Health, Nuclear Safety, and Quality Assurance programs collectively ensure protection of public and worker health and safety and the environment. This is accomplished by empowering and holding accountable managers, employees and contractors to prioritize health, safety and environmental

  1. 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).

  2. Order Module--DOE O 452.1D, NUCLEAR EXPLOSIVE AND WEAPON SURETY PROGRAM, DOE O 452.2D, NUCLEAR EXPLOSIVE SAFETY

    Office of Energy Efficiency and Renewable Energy (EERE)

    "To prevent accidents and inadvertent or unauthorized use of U.S. nuclear weapons and nuclear explosives. In conjunction with the Department of Defense (DoD), to protect the public health and...

  3. 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.

  4. Office Of Nuclear Energy

    Energy Savers [EERE]

    (A Method for Quantifying the Dependability 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 2 Project Overview n Goal, and Objectives Develop measures and methods to assess dependability attributes early and throughout the life-cycle process of software development n Participants * University PI: Dr. Carol Smidts, The Ohio State University (Started February 1,

  5. Nuclear Safety Regulatory Framework

    Office of Environmental Management (EM)

    ... DOE Policy 226.1B, Department of Energy Oversight Policy (April 25, 2011), establishes ... For Further Information Visit: * DOE P 226.1B Department of Energy Oversight Policy (April ...

  6. 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...

  7. 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...

  8. 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.

  9. Integrated Safety Management Safety Culture Resources | Department of

    Energy Savers [EERE]

    Energy Safety Culture Resources Integrated Safety Management Safety Culture Resources A collection of resources available in implementing ISM safety culture activities Safety from the Operator's Perspective: We are All in This Together (2005) Transcript, Keeping the Edge: Enhancing Performance Through Managing Culture (2003), Edgar H. Schein, Ph.D. Proceedings of the Advisory Committee on Reactor Safeguards Safety Culture Workshop (2003) Safety Culture in Nuclear Installations: Guidance for

  10. nuclear

    National Nuclear Security Administration (NNSA)

    2%2A en U.S-, Japan Exchange Best Practices on Nuclear Emergency Response http:nnsa.energy.govmediaroompressreleasesu.s-japan-exchange-best-practices-nuclear-emergency-respon...

  11. Nuclear Security Enterprise | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Study options for ensuring the safety, security, and reliability of nuclear warheads on a ... required to ensure the long-term safety, security, and reliability of the nuclear arsenal.

  12. Nuclear Energy Institute

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

    more understanding of safety margins while addressing ... output of existing nuclear reactors by focusing on important ... safety & feasibility 1% power derating translates to ...

  13. 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.

  14. CSB Investigations and Safety Culture | Department of Energy

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

    from occurring again. Analysis of Safety Systems CSB Investigations and Safety Culture More Documents & Publications Nuclear Safety Workshop Summary Operating Experience...

  15. Senior Technical Safety Manager Qualification Program Self-Assessment...

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

    Senior Technical Safety Manager Qualification Program Self-Assessment - Chief of Nuclear Safety Senior Technical Safety Manager Qualification Program Self-Assessment - Chief of...

  16. NNSA issues Preliminary Notice of Violation to National Security Technologies, LLC, for Nuclear Safety Violations, Fact Sheet

    Broader source: Energy.gov [DOE]

    On August 22, 2011, the National Nuclear Security Administration (NNSA) issued a Preliminary Notice of Violation (PNOV) to National Security Technologies, LLC (NSTec) for violations of Department...

  17. Celebrating Innovation with National Nuclear Science Week | Department...

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

    Did you know that nuclear science is used in archeology, food safety and nuclear medicine? ... tools in the areas of nuclear physics, nuclear chemistry and nuclear engineering. ...

  18. Department of Energy Cites Nuclear Waste Partnership, LLC and...

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

    Nuclear Waste Partnership, LLC and Los Alamos National Security, LLC for Violations Related to Worker Safety and Health and Nuclear Safety Department of Energy Cites Nuclear Waste ...

  19. Nuclear Energy Systems Lab

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

    Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management ...

  20. Advanced Nuclear Energy

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

    Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management ...

  1. 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.

  2. 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.

  3. Pantex Plant final safety analysis report, Zone 4 magazines. Staging or interim storage for nuclear weapons and components: Issue D

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    This Safety Analysis Report (SAR) contains a detailed description and evaluation of the significant environmental, safety, and health (ES&H) issues associated with the operations of the Pantex Plant modified-Richmond and steel arch construction (SAC) magazines in Zone 4. It provides (1) an overall description of the magazines, the Pantex Plant, and its surroundings; (2) a systematic evaluations of the hazards that could occur as a result of the operations performed in these magazines; (3) descriptions and analyses of the adequacy of the measures taken to eliminate, control, or mitigate the identified hazards; and (4) analyses of potential accidents and their associated risks.

  4. Technical Basis for U. S. Department of Energy Nuclear Safety Policy, DOE Policy 420.1, 7/11

    Broader source: Energy.gov [DOE]

    It is the policy of the Department of Energy (DOE) to design, construct, operate anddecommission its nuclear facilities in a manner that ensures adequate protection ofworkers, the public, and the...

  5. 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.

  6. Worker Safety and Health Program for DOE (Including the National Nuclear Security Administration) Federal and Contractor Employees

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

    2011-10-20

    This document was developed to assist the Department of Energy (DOE or the Department) Federal and contractor employees in effectively developing, managing, and implementing a worker safety and health program. Cancels DOE G 440.1-1A and DOE G 440.1-8. Adm Chg 1, dated 3-22-13.

  7. Worker Safety and Health Program for DOE (Including the National Nuclear Security Administration) Federal and Contractor Employees

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

    2011-10-20

    This document was developed to assist the Department of Energy (DOE or the Department) Federal and contractor employees in effectively developing, managing, and implementing a worker safety and health program. Supersedes DOE G 440.1-1A and DOE G 440.1-8.

  8. CRAD, Criteria and Guidelines For the Assessment of Safety System Software and Firmware at Defense Nuclear Facilities

    Broader source: Energy.gov [DOE]

    The purpose and scope of this CRAD is to provide a set of consistent assessment criteria and guidelines for the assessment of safety system software and firmware that performs an SC or SS function, as described in the Background section. The scope of the assessment, henceforth, is called "I&C software."

  9. Nuclear Energy!

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

    Nuclear Energy Technical Assistance Nuclear Energy Technical Assistance "The United States will continue to promote the safe and secure use of nuclear power worldwide through a variety of bilateral and multilateral engagements. For example, the U.S. Nuclear Regulatory Commission advises international partners on safety and regulatory best practices, and the Department of Energy works with international partners on research and development, nuclear waste and storage, training, regulations,

  10. Summary of the contractor information exchange meeting for improving the safety of Soviet-Designed Nuclear Power Plants, February 19, 1997

    SciTech Connect (OSTI)

    1997-04-01

    This report summarizes a meeting held on February 19, 1997, in Washington, D.C. The meeting was held primarily to exchange information among the contractors involved in the U.S. Department of Energy`s efforts to improve the safety of Soviet-designed nuclear power plants. Previous meetings have been held on December 5-6, 1995, and May 22, 1996. The meetings are sponsored by the U.S. Department of Energy and coordinated by the Pacific Northwest National Laboratory. The U.S. Department of Energy works with countries to increase the level of safety at 63 Soviet-designed nuclear reactors operating in Armenia, Bulgaria, the Czech Republic, Hungary, Lithuania, Russia, Slovakia, and Ukraine. The work is implemented largely by commercial companies and individuals who provide technologies and services to the countries with Soviet-designed nuclear power plants. Attending the meeting were 71 representatives of commercial contractors, the U.S. Department of Energy, the U.S. Department of State, national laboratories, and other federal agencies. The presentations and discussions that occurred during the exchange are summarized in this report. While this report captures the general presentation and discussion points covered at the meeting, it is not a verbatim, inclusive record. To make the report useful, information presented at the meeting has been expanded to clarify issues, respond to attendees` requests, or place discussion points in a broader programmatic context. Appendixes A through F contain the meeting agenda, list of attendees, copies of presentation visuals and handouts, the Strategy Document discussed at the meeting, and a summary of attendees` post-meeting evaluation comments. As with past information exchanges, the participants found this meeting valuable and useful. In response to the participant`s requests, a fourth information exchange will be held later in 1997.

  11. NA 50 - Associate Administrator for Safety, Infrastructure and Operations

    National Nuclear Security Administration (NNSA)

    | National Nuclear Security Administration 50 - Associate Administrator for Safety, Infrastructure and Operations FY15

  12. Design of Safety Significant Safety Instrumented Systems Used...

    Office of Environmental Management (EM)

    (SIS) that include both analog and digital control systems are widely used in many industries, including in commercial nuclear power plants, for safety-related applications. ...

  13. DOE Strengthens Rules Governing Worker Safety | Department of...

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

    ... DOE Cites Washington TRU Solutions for Nuclear Safety Violations DOE Cites Safety and Ecology Corp. for Violating Nuclear Safety Rules DOE Cites CH2M Hill Hanford for Violating ...

  14. Request to Cancel DOE M 452.2-2 Admin Chg 1, Nuclear Explosive Safety Evaluation Process

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

    2014-04-21

    NNSA has completed a top down review of four directives that address nuclear explosives operations and is developing a revision to clarify the purpose of conducting NES evaluations, the nature and categorization of findings, and timing of reviews for ongoing operations.

  15. YALINA facility a sub-critical Accelerator- Driven System (ADS) for nuclear energy research facility description and an overview of the research program (1997-2008).

    SciTech Connect (OSTI)

    Gohar, Y.; Smith, D. L.; Nuclear Engineering Division

    2010-04-28

    The YALINA facility is a zero-power, sub-critical assembly driven by a conventional neutron generator. It was conceived, constructed, and put into operation at the Radiation Physics and Chemistry Problems Institute of the National Academy of Sciences of Belarus located in Minsk-Sosny, Belarus. This facility was conceived for the purpose of investigating the static and dynamic neutronics properties of accelerator driven sub-critical systems, and to serve as a neutron source for investigating the properties of nuclear reactions, in particular transmutation reactions involving minor-actinide nuclei. This report provides a detailed description of this facility and documents the progress of research carried out there during a period of approximately a decade since the facility was conceived and built until the end of 2008. During its history of development and operation to date (1997-2008), the YALINA facility has hosted several foreign groups that worked with the resident staff as collaborators. The participation of Argonne National Laboratory in the YALINA research programs commenced in 2005. For obvious reasons, special emphasis is placed in this report on the work at YALINA facility that has involved Argonne's participation. Attention is given here to the experimental program at YALINA facility as well as to analytical investigations aimed at validating codes and computational procedures and at providing a better understanding of the physics and operational behavior of the YALINA facility in particular, and ADS systems in general, during the period 1997-2008.

  16. 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...

  17. Delivering safety

    SciTech Connect (OSTI)

    Baldwin, N.D.; Spooner, K.G.; Walkden, P.

    2007-07-01

    In the United Kingdom there have been significant recent changes to the management of civil nuclear liabilities. With the formation in April 2005 of the Nuclear Decommissioning Authority (NDA), ownership of the civil nuclear licensed sites in the UK, including the Magnox Reactor Stations, passed to this new organisation. The NDAs mission is to seek acceleration of the nuclear clean up programme and deliver increased value for money and, consequently, are driving their contractors to seek more innovative ways of performing work. British Nuclear Group manages the UK Magnox stations under contract to the NDA. This paper summarises the approach being taken within its Reactor Sites business to work with suppliers to enhance working arrangements at sites, improve the delivery of decommissioning programmes and deliver improvements in safety and environmental performance. The UK Magnox stations are 1. generation gas-graphite reactors, constructed in the 1950's and 1960's. Two stations are currently still operating, three are shut-down undergoing defueling and the other five are being decommissioned. Despite the distractions of industry restructuring, an uncompromising policy of demanding improved performance in conjunction with improved safety and environmental standards has been adopted. Over the past 5 years, this policy has resulted in step-changes in performance at Reactor Sites, with increased electrical output and accelerated defueling and decommissioning. The improvements in performance have been mirrored by improvements in safety (DACR of 0 at 5 sites); environmental standards (reductions in energy and water consumption, increased waste recycling) and the overall health of the workforce (20% reduction in sickness absence). These achievements have, in turn, been recognised by external bodies, resulting in several awards, including: the world's first ISRS and IERS level 10 awards (Sizewell, 2006), the NUMEX plant maintenance award (Bradwell, 2006), numerous RoSPA awards at site and sector level and nomination, at Company level, for the RoSPA George Earle trophy for outstanding performance in Health and Safety (Reactor Sites, 2006). After 'setting the scene' and describing the challenges that the company has had to respond to, the paper explains how these improvements have been delivered. Specifically it explains the process that has been followed and the parts played by sites and suppliers to deliver improved performance. With the experience of already having transitioned several Magnox stations from operations to defueling and then to decommissioning, the paper describes the valuable experience that has been gained in achieving an optimum change process and maintaining momentum. (authors)

  18. Facility Safety Policy, Guidance & Reports | Department of Energy

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

    Facility Safety Policy, Guidance & Reports Facility Safety Policy, Guidance & Reports The Office of Nuclear Facility Safety Programs within the U.S. Department of Energy's (DOE) Environment, Health, Safety and Security organization, establishes nuclear safety requirements related to safety management programs that are essential to the safety of DOE nuclear facilities. In addition, establishes requirements for facility design and operation for facility-wide hazards that are not unique to

  19. Summary and bibliography of safety-related events at boiling-water nuclear power plants as reported in 1980

    SciTech Connect (OSTI)

    McCormack, K.E.; Gallaher, R.B.

    1982-03-01

    This document presents a bibliography that contains 100-word abstracts of event reports submitted to the US Nuclear Regulatory Commission concerning operational events that occurred at boiling-water-reactor nuclear power plants in 1980. The 1547 abstracts included on microfiche in this bibliography describe incidents, failures, and design or construction deficiencies that were experienced at the facilities. These abstracts are arranged alphabetically by reactor name and then chronologically for each reactor. Full-size keyword and permuted-title indexes to facilitate location of individual abstracts are provided following the text. Tables that summarize the information contained in the bibliography are also provided. The information in the tables includes a listing of the equipment items involved in the reported events and the associated number of reports for each item. Similar information is given for the various kinds of instrumentation and systems, causes of failures, deficiencies noted, and the time of occurrence (i.e., during refueling, operation, testing, or construction).

  20. Risk and Safety Assessment

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

    and Safety Assessment - Sandia Energy Energy Search Icon Sandia Home Locations Contact Us Employee Locator Energy & Climate Secure & Sustainable Energy Future Stationary Power Energy Conversion Efficiency Solar Energy Wind Energy Water Power Supercritical CO2 Geothermal Natural Gas Safety, Security & Resilience of the Energy Infrastructure Energy Storage Nuclear Power & Engineering Grid Modernization Battery Testing Nuclear Fuel Cycle Defense Waste Management Programs Advanced