Sample records for nuclear criticality safety

  1. Nuclear Engineering Nuclear Criticality Safety

    E-Print Network [OSTI]

    Kemner, Ken

    development, Nuclear Operations Division (NOD) waste management and storage activities and other laboratoryNuclear Engineering Nuclear Criticality Safety The Nuclear Engineering Division (NE) of Argonne National Laboratory is experienced in performing criticality safety and shielding evaluations for nuclear

  2. Autoclave nuclear criticality safety analysis

    SciTech Connect (OSTI)

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

    1991-12-31T23:59:59.000Z

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

  3. Elements of a nuclear criticality safety program

    SciTech Connect (OSTI)

    Hopper, C.M.

    1995-07-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    WIPP-016, Rev. 0 Nuclear Criticality Safety Evaluation for Contact-Handled Transuranic Waste/2008 Guidance (if applicable): _______________________ #12;NUCLEAR CRITICALITY SAFETY EVALUATION FOR CONTACT, directors, employees, agents, consultants or personal services contractors. #12;NUCLEAR CRITICALITY SAFETY

  5. Nuclear criticality safety department training implementation

    SciTech Connect (OSTI)

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

    1996-09-06T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Not Available

    1994-02-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    at the Waste Isolation Pilot Plant #12;SUMMARY OF NUCLEAR CRITICALITY SAFETY EVALUATION FOR SHIELDED CONTAINERS PLANT WIPP-025, REV. 0 AUGUST 2009 Summary of Nuclear Criticality Safety Evaluation for Shielded ISOLATION PILOT PLANT WIPP-025, REV. 0 AUGUST 2009 ES-1 Executive Summary This report summarizes the nuclear

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

    SciTech Connect (OSTI)

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

    2009-11-01T23:59:59.000Z

    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.

  9. Nuclear criticality safety engineer qualification program utilizing SAT

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Sanchez, R.G. [comp.

    1994-01-01T23:59:59.000Z

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

  11. Nuclear criticality safety tools in the Chernobyl-4 accident analysis

    SciTech Connect (OSTI)

    Landeyro, P.A.

    1988-01-01T23:59:59.000Z

    The collaboration with the Italian Safety Authority (DISP), started in July 1986, has the aim of studying, from a neutronic point of view, the possible initiator event and the accident dynamics in unit four of the Chernobly nuclear power plant. This report was produced within the framework of that collaboration. A main condition of the present work was making use of standard calculational methods employed in nuclear criticality safety analysis. This means that the neutron multiplication factor calculation should be made with the modules and the cross-section libraries of the SCALE system or in any case with some KENO IV version and the burnup calculation with the ORIGEN code.

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

    E-Print Network [OSTI]

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

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

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

    2007-02-07T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Hopper, Calvin Mitchell [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1993-09-20T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-01-01T23:59:59.000Z

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

  17. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    SciTech Connect (OSTI)

    Not Available

    1993-11-01T23:59:59.000Z

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24.

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

    SciTech Connect (OSTI)

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

    2007-04-30T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    2006-01-01T23:59:59.000Z

    241 Pu, etc. ). To prevent nuclear criticality in spent fuelto enhance criticality safety for spent nuclear fuel inSpent Nuclear Fuel (SNF) Container to Enhance Criticality

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

    SciTech Connect (OSTI)

    Huffer, J.E. [Parallax, Inc., Atlanta, GA (United States)

    1997-04-01T23:59:59.000Z

    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..

  1. Assessment of safety-critical software in nuclear power plants

    SciTech Connect (OSTI)

    Parnas, D.L.; Madey, J. [McMaster Univ., Hamilton, Ontario (Canada); Asmis, G.J.K. [Atomic Energy Control Board, Ottawa (Canada)

    1991-04-01T23:59:59.000Z

    This article outlines an approach in the design, documentation, and evaluation of computer systems. This allows the use of software in many safety-critical applications because it enables the systematic comparison of the program behavior with the engineering specifications of the computer system. Many of the ideas in this article have been used by the Atomic Energy Control Board of Canada in its safety assessment of the software for the shutdown systems of the Darlington Station. The four main elements of this approach follow: (1) Formal Documentation of Software Requirements: Most of the details of a complex environment can be ignored by system implementers and reviewers if they are given a complete and precise statement of the behavioral requirements for the computer system. We describe five mathematical relations that specify the requirements for the software in a computerized control system. (2) Design and Documentation of the Module Structure: Complexity caused by interactions between separately written components can be reduced by applying Data Abstraction, Abstract Data Types, and Object-Oriented Programming if the interfaces are precisely and completely documented. (3) Program Function Documentation: Software executions are lengthy sequences of state changes described by algorithms. The effects of these executive sequences can be precisely specified documented with tabular presentations of the program functions. Also, large programs can be decomposed and presented at a collection of well-documented smaller programs. (4) Tripod Approach to Assessment: There are three basic approaches to the assessment of complex software products: (i) testing, (ii) systematic inspection, and (iii) certification of people and processes. Assessment of a complex system cannot depend on any one of these alone. The approach used on the Darlington shutdown software, which included systematic inspection as well as planned and statistically designed random testing, is outlined.

  2. Nuclear criticality safety in D and D operations: a Los Alamos experience

    SciTech Connect (OSTI)

    Schlesser, J.A.

    1996-12-31T23:59:59.000Z

    Decommissioning operations at the Los Alamos National Laboratory require the interaction of several disciplines so that the effort to D&D radiological facilities can proceed unencumbered, on schedule, and within budget. Although playing a minor role, the Laboratory`s Nuclear Criticality Safety Group has provided criticality safety guidance to one such D&D team efficiently and cost-effectively. During the first major D&D effort at Los Alamos, a total of about 6 kilograms of uranium [U(93)] was recovered from a facility thought to contain only tens of grams.

  3. The Development, Content, Design, and Conduct of the 2011 Piloted US DOE Nuclear Criticality Safety Program Criticality Safety Engineering Training and Education Project

    SciTech Connect (OSTI)

    Hopper, Calvin Mitchell [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    In May 1973 the University of New Mexico conducted the first nationwide criticality safety training and education week-long short course for nuclear criticality safety engineers. Subsequent to that course, the Los Alamos Critical Experiments Facility (LACEF) developed very successful 'hands-on' subcritical and critical training programs for operators, supervisors, and engineering staff. Since the inception of the US Department of Energy (DOE) Nuclear Criticality Technology and Safety Project (NCT&SP) in 1983, the DOE has stimulated contractor facilities and laboratories to collaborate in the furthering of nuclear criticality as a discipline. That effort included the education and training of nuclear criticality safety engineers (NCSEs). In 1985 a textbook was written that established a path toward formalizing education and training for NCSEs. Though the NCT&SP went through a brief hiatus from 1990 to 1992, other DOE-supported programs were evolving to the benefit of NCSE training and education. In 1993 the DOE established a Nuclear Criticality Safety Program (NCSP) and undertook a comprehensive development effort to expand the extant LACEF 'hands-on' course specifically for the education and training of NCSEs. That successful education and training was interrupted in 2006 for the closing of the LACEF and the accompanying movement of materials and critical experiment machines to the Nevada Test Site. Prior to that closing, the Lawrence Livermore National Laboratory (LLNL) was commissioned by the US DOE NCSP to establish an independent hands-on NCSE subcritical education and training course. The course provided an interim transition for the establishment of a reinvigorated and expanded two-week NCSE education and training program in 2011. The 2011 piloted two-week course was coordinated by the Oak Ridge National Laboratory (ORNL) and jointly conducted by the Los Alamos National Laboratory (LANL) classroom education and facility training, the Sandia National Laboratory (SNL) hands-on criticality experiments training, and the US DOE National Criticality Experiment Research Center (NCERC) hands-on criticality experiments training that is jointly supported by LLNL and LANL and located at the Nevada National Security Site (NNSS) This paper provides the description of the bases, content, and conduct of the piloted, and future US DOE NCSP Criticality Safety Engineer Training and Education Project.

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

    SciTech Connect (OSTI)

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

    2011-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1993-12-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1993-08-16T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Marshall, William BJ J [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

    2012-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    J. Blair Briggs; Anatoly Tsibulya; Yevgeniy Rozhikhin

    2012-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2013-10-01T23:59:59.000Z

    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.

  11. Submersion Criticality Safety Analysis of Tungsten-Based Fuel for Nuclear Power and Propulsion Applications

    SciTech Connect (OSTI)

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

    2014-07-01T23:59:59.000Z

    The Center for Space Nuclear Research (CSNR) is developing tungsten-encapsulated fuels for space nuclear applications. Aims to develop NTP fuels that are; Affordable Low impact on production and testing environment Producible on a large scale over suitable time period Higher-performance compared to previous graphite NTP fuel elements Space nuclear reactors remain subcritical before and during launch, and do not go critical until required by its mission. A properly designed reactor will remain subcritical in any launch abort scenario, where the reactor falls back to Earth and becomes submerged in terrestrial material. Submersion increases neutron reflection and thermalizes the neutrons, which typically increases the reactivity of the core. This effect is usually very significant for fast-spectrum reactors. This research provided a submersion criticality safety analysis for a representative tungsten/uranium oxide fueled reactor. Determine the submersion behavior of a reactor fueled by tungsten-based fuel. Considered fuel compositions with varying: Rhenium content (wt% rhenium in tungsten) Fuel loading fractions (UO2 vol%)

  12. Additional Studies of the Criticality Safety of Failed Used Nuclear Fuel

    SciTech Connect (OSTI)

    Marshall, William BJ J [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

    2013-01-01T23:59:59.000Z

    Commercial used nuclear fuel (UNF) in the United States is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high-burnup values (>45 GWd/t) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF is not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories and specific configurations were evaluated to understand trends and quantify the consequences of worst-case potential reconfiguration progressions. These results will be summarized here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g., >20% keff). It can be concluded that the consequences of credible fuel failure configurations from ES or transportation following ES are manageable (e.g., <5% keff). The current work expands on these efforts and examines some modified scenarios and modified approaches to investigate the effectiveness of some techniques for reducing the calculated increase in keff. The areas included here are more realistic modeling of some assembly types and the effect of reconfiguration of some assemblies in the storage and transportation canister.

  13. Criticality Safety Evaluation of a LLNL Training Assembly for Criticality Safety (TACS)

    SciTech Connect (OSTI)

    Heinrichs, D P

    2006-06-26T23:59:59.000Z

    Hands-on experimental training in the physical behavior of multiplying systems is one of ten key areas of training required for practitioners to become qualified in the discipline of criticality safety as identified in DOE-STD-1135-99, ''Guidance for Nuclear Criticality Safety Engineer Training and Qualification''. This document is a criticality safety evaluation of the training activities (or operations) associated with HS-3200, ''Laboratory Class for Criticality Safety''. These activities utilize the Training Assembly for Criticality Safety (TACS). The original intent of HS-3200 was to provide LLNL fissile material handlers with a practical hands-on experience as a supplement to the academic training they receive biennially in HS-3100, ''Fundamentals of Criticality Safety'', as required by ANSI/ANS-8.20-1991, ''Nuclear Criticality Safety Training''. HS-3200 is to be enhanced to also address the training needs of nuclear criticality safety professionals under the auspices of the NNSA Nuclear Criticality Safety Program.

  14. Stochastic sampling method with MCNPX for nuclear data uncertainty propagation in criticality safety applications

    SciTech Connect (OSTI)

    Zhu, T.; Vasiliev, A.; Wieselquist, W.; Ferroukhi, H. [Paul Scherrer Institut, 5232 Villigen (Switzerland)

    2012-07-01T23:59:59.000Z

    In the domain of criticality safety, the efficient propagation of uncertainty in nuclear data to uncertainty in k{sub eff} is an important area of current research. In this paper, a method based on stochastic sampling is presented for uncertainty propagation in MCNPX calculations. To that aim, the nuclear data (i.e. cross sections) are assumed to have a multivariate normal distribution and simple random sampling is performed following this presumed probability distribution. A verification of the developed stochastic sampling procedure with MCNPX is then conducted using the {sup 239}Pu Jezebel experiment as well as the PB-2 BWR and TMI-1 PWR pin cell models from the Uncertainty Analysis in Modeling (UAM) exercises. For the Jezebel case, it is found that the developed stochastic sampling approach predicts similar k{sub eff} uncertainties compared to conventional sensitivity and uncertainty methods. For the UAM models, slightly lower uncertainties are obtained when comparing to existing preliminary results. Further details of these verification studies are discussed and directions for future work are outlined. (authors)

  15. 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.

    1982-10-21T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Frost, R.L.

    1999-02-26T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Sobes, Vladimir [ORNL] [ORNL; Leal, Luiz C [ORNL] [ORNL; Guber, Klaus H [ORNL] [ORNL; Forget, Benoit [Massachusetts Institute of Technology (MIT)] [Massachusetts Institute of Technology (MIT); Kopecky, S. [EC-JRC-IRMM, Geel, Belgium] [EC-JRC-IRMM, Geel, Belgium; Schillebeeckx, P. [EC-JRC-IRMM, Geel, Belgium] [EC-JRC-IRMM, Geel, Belgium; Siegler, P. [EC-JRC-IRMM, Geel, Belgium] [EC-JRC-IRMM, Geel, Belgium

    2014-01-01T23:59:59.000Z

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

  18. CRITICALITY SAFETY TRAINING AT FLUOR HANFORD (FH)

    SciTech Connect (OSTI)

    TOFFER, H.

    2005-05-02T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Danon, Yaron

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

  20. Nuclear Safety Regulatory Framework

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

    overall Nuclear Safety Policy & ESH Goals Safety Basis Review and Approval In the DOE governance model, contractors responsible for the facility develop the safety basis and...

  1. Nuclear criticality safety bounding analysis for the in-tank-precipitation (ITP) process, impacted by fissile isotopic weight fractions

    SciTech Connect (OSTI)

    Bess, C.E.

    1994-04-22T23:59:59.000Z

    The In-Tank Precipitation process (ITP) receives High Level Waste (HLW) supernatant liquid containing radionuclides in waste processing tank 48H. Sodium tetraphenylborate, NaTPB, and monosodium titanate (MST), NaTi{sub 2}O{sub 5}H, are added for removal of radioactive Cs and Sr, respectively. In addition to removal of radio-strontium, MST will also remove plutonium and uranium. The majority of the feed solutions to ITP will come from the dissolution of supernate that had been concentrated by evaporation to a crystallized salt form, commonly referred to as saltcake. The concern for criticality safety arises from the adsorption of U and Pt onto MST. If sufficient mass and optimum conditions are achieved then criticality is credible. The concentration of u and Pt from solution into the smaller volume of precipitate represents a concern for criticality. This report supplements WSRC-TR-93-171, Nuclear Criticality Safety Bounding Analysis For The In-Tank-Precipitation (ITP) Process. Criticality safety in ITP can be analyzed by two bounding conditions: (1) the minimum safe ratio of MST to fissionable material and (2) the maximum fissionable material adsorption capacity of the MST. Calculations have provided the first bounding condition and experimental analysis has established the second. This report combines these conditions with canyon facility data to evaluate the potential for criticality in the ITP process due to the adsorption of the fissionable material from solution. In addition, this report analyzes the potential impact of increased U loading onto MST. Results of this analysis demonstrate a greater safety margin for ITP operations than the previous analysis. This report further demonstrates that the potential for criticality in the ITP process due to adsorption of fissionable material by MST is not credible.

  2. Tank farms criticality safety manual

    SciTech Connect (OSTI)

    FORT, L.A.

    2003-03-27T23:59:59.000Z

    This document defines the Tank Farms Contractor (TFC) criticality safety program, as required by Title 10 Code of Federal Regulations (CFR), Subpart 830.204(b)(6), ''Documented Safety Analysis'' (10 CFR 830.204 (b)(6)), and US Department of Energy (DOE) 0 420.1A, Facility Safety, Section 4.3, ''Criticality Safety.'' In addition, this document contains certain best management practices, adopted by TFC management based on successful Hanford Site facility practices. Requirements in this manual are based on the contractor requirements document (CRD) found in Attachment 2 of DOE 0 420.1A, Section 4.3, ''Nuclear Criticality Safety,'' and the cited revisions of applicable standards published jointly by the American National Standards Institute (ANSI) and the American Nuclear Society (ANS) as listed in Appendix A. As an informational device, requirements directly imposed by the CRD or ANSI/ANS Standards are shown in boldface. Requirements developed as best management practices through experience and maintained consistent with Hanford Site practice are shown in italics. Recommendations and explanatory material are provided in plain type.

  3. 2011 Annual Criticality Safety Program Performance Summary

    SciTech Connect (OSTI)

    Andrea Hoffman

    2011-12-01T23:59:59.000Z

    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

  4. Office of Nuclear Safety

    Broader source: Energy.gov [DOE]

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

  5. Lecture notes for criticality safety

    SciTech Connect (OSTI)

    Fullwood, R.

    1992-03-01T23:59:59.000Z

    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.

  6. Nuclear Safety Research and Development...

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

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

  7. Nuclear Multifragmentation Critical Exponents

    E-Print Network [OSTI]

    Wolfgang Bauer; William Friedman

    1994-11-14T23:59:59.000Z

    We show that the critical exponents of nuclear multi-fragmentation have not been determined conclusively yet.

  8. Nuclear Explosive Safety Manual

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

    2009-04-14T23:59:59.000Z

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

  9. Nuclear Explosive Safety

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

    2015-01-26T23:59:59.000Z

    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).

  10. Nuclear Explosive Safety

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

    2014-07-10T23:59:59.000Z

    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).

  11. Nuclear Safety (Pennsylvania)

    Broader source: Energy.gov [DOE]

    The Nuclear Safety Division conducts a comprehensive nuclear power plant oversight review program of the nine reactors at the five nuclear power sites in Pennsylvania. It also monitors the...

  12. Reference handbook: Nuclear criticality

    SciTech Connect (OSTI)

    Not Available

    1991-12-06T23:59:59.000Z

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

  13. Promulgating Nuclear Safety Requirements

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

    1996-05-15T23:59:59.000Z

    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.

  14. Nuclear Explosive Safety

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

    2009-04-14T23:59:59.000Z

    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

  15. Nuclear Explosive Safety

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

    2009-04-14T23:59:59.000Z

    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.

  16. 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.

  17. Nuclear Explosive Safety

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

    2006-06-12T23:59:59.000Z

    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.

  18. CRITICALITY SAFETY CONTROLS AND THE SAFETY BASIS AT PFP

    SciTech Connect (OSTI)

    Kessler, S

    2009-04-21T23:59:59.000Z

    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

  19. Nuclear Explosive Safety

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

    2006-06-12T23:59:59.000Z

    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.

  20. INTEGRATION OF ADAPTIVE FILE ASSIGNMENT INTO DISTRIBUTED SAFETY-CRITICAL SYSTEMS

    E-Print Network [OSTI]

    Wedde, Horst F.

    . In safety-critical systems (such as nuclear power plants, autonomous cooperation of robots in Outer Space-Critical / Safety-critical Systems, Functional/ Operational Integration 1. INTRODUCTION Safety-critical systems that are typically unpredictable, a very high amount of adaptability of system functions is demanded. safety

  1. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2005-04-07T23:59:59.000Z

    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].

  2. 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.

  3. Nuclear Safety Information Dashboard | Department of Energy

    Office of Environmental Management (EM)

    Nuclear Safety Information Dashboard Nuclear Safety Information Dashboard The Nuclear Safety Information (NSI) Dashboard provides a new user interface to the Occurrence Reporting...

  4. Nuclear Explosive Safety Manual - DOE Directives, Delegations...

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

    1A Admin Chg 1, Nuclear Explosive Safety Manual by Carl Sykes Functional areas: Administrative Change, Defense Nuclear Facility Safety and Health Requirement, Nuclear Safety,...

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

    SciTech Connect (OSTI)

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

    1991-12-31T23:59:59.000Z

    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.

  6. Nuclear Explosive Safety Manual

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

    2009-04-14T23:59:59.000Z

    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.

  7. Nuclear Explosive Safety Evaluation Processes

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

    2009-04-14T23:59:59.000Z

    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.

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

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

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

  9. Nuclear explosive safety study process

    SciTech Connect (OSTI)

    NONE

    1997-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Jee, Eunkyoung

    of nuclear power plants. The various techniques applied to a safety analysis on the structures and systemsSafety Activities on Safety-Critical Software for Reactor Protection System Gee-Yong Park1 , Kee Institute, P.O.Box 105 Yuseong, Daejeon, 305-353 KOREA 2: Korea Advanced Institute of Science and Technology

  11. Facility Safety

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

    1996-10-24T23:59:59.000Z

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

  12. Facility Safety

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

    1995-11-16T23:59:59.000Z

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

  13. Criticality Safety | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613PortsmouthBartlesvilleAbout » Contact Us ContactPracticesWinter (PartCriticality Safety

  14. 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 [Cabrera Services (United States); Matthews, Brian [Nuclear Safety Associates (United States)

    2012-07-01T23:59:59.000Z

    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)

  15. Nuclear Reactions and Reactor Safety

    E-Print Network [OSTI]

    Onuchic, José

    Nuclear Reactions and Reactor Safety DO NOT LICK We haven't entirely nailed down what element nuclear chain reaction, 1938 #12;Nuclear Chain Reactions Do nuclear chain reactions lead to runaway explosions? or ? -Controlled nuclear chain reactions possible: control energy release/sec -> More

  16. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01T23:59:59.000Z

    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.

  17. Nuclear Safety Regulatory Framework

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin of Contamination in ManyDepartment of Energy NorthBDepartment of Energy Nuclear Safety

  18. CRAD, Facility Safety- Nuclear Facility Safety Basis

    Broader source: Energy.gov [DOE]

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

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

    E-Print Network [OSTI]

    A Domain-Specific Safety Analysis for Digital Nuclear Plant Protection Systems Sanghyun Yoon through safety analy- sis is strongly mandated for safety-critical systems. Nuclear plant protection. INTRODUCTION Safety-critical systems (e.g. nuclear power plants and air- planes) require rigorous quality

  20. Nuclear Explosive Safety Evaluation Processes

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

    2009-04-14T23:59:59.000Z

    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.

  1. Nuclear Reactor Safety Design Criteria

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

    1993-01-19T23:59:59.000Z

    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. Certified 11-18-10.

  2. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, E.

    1983-08-15T23:59:59.000Z

    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.

  3. SHSD Manager Safety Engineering Group Manager

    E-Print Network [OSTI]

    Safety, Machine Shop Safety, Tier I Program, Traffic Safety S. Moss: Nuclear Criticality Safety G. Shepherd: Explosives Safety, Facility Authorization Basis, Nuclear Safety R. Travis: Readiness Evaluations

  4. Safety Lifecycle for Developing Safety Critical Artificial Neural Networks

    E-Print Network [OSTI]

    Kelly, Tim

    Safety Lifecycle for Developing Safety Critical Artificial Neural Networks Zeshan Kurd, Tim Kelly.kelly}@cs.york.ac.uk Abstract. Artificial neural networks are employed in many areas of industry such as medicine and defence a safety lifecycle for artificial neural networks. The lifecycle fo- cuses on managing behaviour

  5. Nuclear Explosive Safety Evaluation Processes - DOE Directives...

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

    2 Admin Chg 1, Nuclear Explosive Safety Evaluation Processes by Carl Sykes Functional areas: Administrative Change, Defense Nuclear Facility Safety and Health Requirement, Defense...

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

    SciTech Connect (OSTI)

    Koponen, B L; Heinrichs, D

    2009-05-18T23:59:59.000Z

    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.

  7. Critical QCD in Nuclear Collisions

    E-Print Network [OSTI]

    N. G. Antoniou; Y. F. Contoyiannis; F. K. Diakonos; G. Mavromanolakis

    2005-05-20T23:59:59.000Z

    A detailed study of correlated scalars, produced in collisions of nuclei and associated with the $\\sigma$-field fluctuations, $(\\delta \\sigma)^2= $, at the QCD critical point (critical fluctuations), is performed on the basis of a critical event generator (Critical Monte-Carlo) developed in our previous work. The aim of this analysis is to reveal suitable observables of critical QCD in the multiparticle environment of simulated events and select appropriate signatures of the critical point, associated with new and strong effects in nuclear collisions.

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

    E-Print Network [OSTI]

    is on increase in the safety-critical system in recent years. In the safety-critical systems such as Nuclear**, and Sung Deok Cha** Korea Advanced Institute of Science and Technology * Department of Nuclear and QuantumAN INTEGRATED ENVIRONMENT OF S/W SPECIFICATION AND V&V FOR SAFETY-CRITICAL SYSTEMS Abstract

  9. Criticality safety assessment of tank 241-C-106 remediation

    SciTech Connect (OSTI)

    Waltar, A.E., Westinghouse Hanford

    1996-07-19T23:59:59.000Z

    A criticality safety assessment was performed in support of Project 320 for the retrieval of waste from tank 241-C-106 to tank 241-AY-102. The assessment was performed by a multi-disciplined team consisting of expertise covering the range of nuclear engineering, plutonium and nuclear waste chemistry,and physical mixing hydraulics. Technical analysis was performed to evaluate the physical and chemical behavior of fissile material in neutralized Hanford waste as well as modeling of the fluid dynamics for the retrieval activity. The team has not found evidence of any credible mechanism to attain neutronic criticality in either tank and has concluded that a criticality accident is incredible.

  10. Criticality Safety Basics for INL FMHs and CSOs

    SciTech Connect (OSTI)

    V. L. Putman

    2012-04-01T23:59:59.000Z

    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.

  11. 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).

  12. 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.

  13. Proving the Absence of RunTime Errors in SafetyCritical Avionics Code

    E-Print Network [OSTI]

    Cousot, Patrick

    , time­triggered, real­time, safety critical, embedded software as found in earth transportation, nuclearProving the Absence of Run­Time Errors in Safety­Critical Avionics Code Patrick Cousot École is not acceptable in safety and mission crit­ ical applications. An avenue is therefore opened for formal methods

  14. NOVEL PRINCIPLES FOR DEVELOPING AND EVALUATING DISTRIBUTED SAFETY-CRITICAL SYSTEMS

    E-Print Network [OSTI]

    Wedde, Horst F.

    . In safety-critical systems, (such as nuclear power plants, distributed cooperation of autonomous robotsNOVEL PRINCIPLES FOR DEVELOPING AND EVALUATING DISTRIBUTED SAFETY-CRITICAL SYSTEMS Horst F. Wedde, Jon A. Lind Informatik III University of Dortmund 44221 Dortmund / Germany Abstract Safety

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

    SciTech Connect (OSTI)

    Putman, Valerie Lee

    2003-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Mattson, Roger J.

    1989-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

    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)

  18. Nuclear reactor safety heat transfer

    SciTech Connect (OSTI)

    Jones, O.C.

    1982-07-01T23:59:59.000Z

    Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.

  19. Safety of Nuclear Explosive Operations

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

    2001-08-07T23:59:59.000Z

    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.

  20. DISTRIBUTED REAL-TIME TASK MONITORING IN THE SAFETY-CRITICAL SYSTEM MELODY

    E-Print Network [OSTI]

    Wedde, Horst F.

    -critical systems (such as nuclear power plants, distributed cooperation of autonomous robots in Outer Space that are typi- cally unpredictable, a very high amount of adaptability of sys- tem functions is demanded. SafetyDISTRIBUTED REAL-TIME TASK MONITORING IN THE SAFETY-CRITICAL SYSTEM MELODY Horst F. Wedde, Jon A

  1. Safety critical software development qualification

    SciTech Connect (OSTI)

    Marron, J. E. [Invensys Process Systems, 33 Commercial Street, Foxboro, MA 02035 (United States)

    2006-07-01T23:59:59.000Z

    With the increasing use of digital systems in control applications, customers must acquire appropriate expectations for software development and quality assurance procedures. Purchasers and users of digital systems need to understand the benefits to the supplier of effective quality systems. These systems consist not only of procedures but tools that enable automation. Without the use of automation, quality can not be assured. A software and systems quality program starts with the documents you are very familiar with. But these documents must define more than the final system. They must address specific development environment characteristics and testing capabilities. Starting with the RFP, some of the items that should be introduced are Software Configuration Management, regression testing and defect tracking. The digital system customer is in the best position to enforce the use of software and systems quality programs by including them in project requirements as early as the Purchase Order. The customer's understanding of the full scope and implementation of a software quality program is essential to achieving the quality necessary in nuclear projects, and, incidentally, completing those projects on schedule. (authors)

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

    SciTech Connect (OSTI)

    Mavko, B.; Kozuh, M.

    1994-12-31T23:59:59.000Z

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

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

    Office of Environmental Management (EM)

    Information Agreement Between the U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards, and the U.S. Department of Energy, Office of Environment,...

  4. ORNL/TM-2011/450 Criticality Safety Validation of Scale 6.1

    E-Print Network [OSTI]

    Government or any agency thereof. #12;ORNL/TM-2011/450 Reactor and Nuclear Systems Division CriticalityORNL/TM-2011/450 Criticality Safety Validation of Scale 6.1 November 2011 Prepared by W. J) representatives, and International Nuclear Information System (INIS) representatives from the following source

  5. Total safety: A new safety culture to integrate nuclear safety and operational safety

    SciTech Connect (OSTI)

    Saji, G. [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan); Murphy, G.A. [ed.

    1991-07-01T23:59:59.000Z

    The creation of a complete and thorough safety culture is proposed for the purpose of providing additional assurance about nuclear safety and improving the performance of nuclear power plants. The safety philosophy developed a combination of the former hardware-oriented nuclear safety approach and recent operational safety concepts. The improvement of the latter, after TMI-2 and Chernobyl, has been proven very effective in reducing the total risk associated with nuclear power plants. The first part of this article introduces a {open_quotes}total safety{close_quotes} concept. This extends the concept of {open_quotes}nuclear safety{close_quotes} and makes it closer to the public perception of safety. This concept is defined by means of a taxonomy of total safety. The second part of the article shows that total safety can be achieved by integrating it into a modern quality assurance (QA) system since it is tailored to make implementation into a framework of QA easier. The author believes that the outstanding success experienced by various industries as a result of introducing the modern QA system should lead to its application for ensuring the safety and performance of nuclear facilities. 15 refs., 3 figs.

  6. Rocky Flats CAAS System Recalibrated, Retested, and Analyzed to Install in the Criticality Experiments Facility at the Nevada Test Site

    E-Print Network [OSTI]

    2009-01-01T23:59:59.000Z

    sponsorship of the DOE Nuclear Criticality Safety Program.Improved Criticality Alarm System,” Proceedings of Nuclear

  7. Criticality Safety Basics for INL Emergency Responders

    SciTech Connect (OSTI)

    Valerie L. Putman

    2012-08-01T23:59:59.000Z

    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.

  8. TRANSPORTATION CASK RECEIPT/RETURN FACILITY CRITICALITY SAFETY EVALUATIONS

    SciTech Connect (OSTI)

    C.E. Sanders

    2005-04-26T23:59:59.000Z

    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].

  9. A philosophy for space nuclear systems safety

    SciTech Connect (OSTI)

    Marshall, A.C.

    1992-08-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Choi, Jor-Shan; Lee, Chuck; Farmer, Joseph; Day, Dan; Wall, Mark; Saw, Cheng [Lawrence Livermore National Laboratory, Livermore, CA, 94550 (United States); Boussoufi, Moe; Liu, Ben; Egbert, Harold [McClellan Nuclear Radiation Center, McClellan, CA, 95652 (United States); Branagan, Dan [Nanosteel Company, Idaho Falls, ID, 83402 (United States); D'Amato, Andy [Plasma Technology Incorporated, Torrance, CA, 90501 (United States)

    2007-07-01T23:59:59.000Z

    Spent nuclear fuel contains fissionable materials ({sup 235}U, {sup 239}Pu, {sup 241}Pu, etc.). To prevent nuclear criticality in spent fuel storage, transportation, and during disposal, neutron-absorbing materials (or neutron poisons, such as borated stainless steel, Boral{sup TM}, Metamic{sup TM}, Ni-Gd, and others) would have to be applied. The success in demonstrating that the High-Performance Corrosion- Resistant Material (HPCRM){sup [1]} can be thermally applied as coating onto base metal to provide for corrosion resistance for many naval applications raises the interest in applying the HPCRM to USDOE/OCRWM spent fuel management program. The fact that the HPCRM relies on the high content of boron to make the material amorphous - an essential property for corrosion resistance - and that the boron has to be homogeneously distributed in the HPCRM qualify the material to be a neutron poison. (authors)

  11. SCALE 6: Comprehensive Nuclear Safety Analysis Code System

    SciTech Connect (OSTI)

    Bowman, Stephen M [ORNL

    2011-01-01T23:59:59.000Z

    Version 6 of the Standardized Computer Analyses for Licensing Evaluation (SCALE) computer software system developed at Oak Ridge National Laboratory, released in February 2009, contains significant new capabilities and data for nuclear safety analysis and marks an important update for this software package, which is used worldwide. This paper highlights the capabilities of the SCALE system, including continuous-energy flux calculations for processing multigroup problem-dependent cross sections, ENDF/B-VII continuous-energy and multigroup nuclear cross-section data, continuous-energy Monte Carlo criticality safety calculations, Monte Carlo radiation shielding analyses with automated three-dimensional variance reduction techniques, one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations, two- and three-dimensional lattice physics depletion analyses, fast and accurate source terms and decay heat calculations, automated burnup credit analyses with loading curve search, and integrated three-dimensional criticality accident alarm system analyses using coupled Monte Carlo criticality and shielding calculations.

  12. Anomalies of Nuclear Criticality, Revision 6

    SciTech Connect (OSTI)

    Clayton, E. D.; Prichard, Andrew W.; Durst, Bonita E.; Erickson, David; Puigh, Raymond J.

    2010-02-19T23:59:59.000Z

    This report is revision 6 of the Anomalies of Nuclear Criticality. This report is required reading for the training of criticality professionals in many organizations both nationally and internationally. This report describes many different classes of nuclear criticality anomalies that are different than expected.

  13. Systems Issues in Nuclear Reactor Safety

    E-Print Network [OSTI]

    de Weck, Olivier L.

    Systems Issues in Nuclear Reactor Safety Commissioner George ApostolakisCommissioner George Apostolakis U.S. Nuclear Regulatory Commission CmrApostolakis@nrc.gov MIT SDM Conference on Systems Thinking, source, and special nuclear materials to ensure adequate protection of public health and safety, 3

  14. Validation of Criticality Safety Calculations with SCALE 6.2

    SciTech Connect (OSTI)

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

    2013-01-01T23:59:59.000Z

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

  15. ME 379M-Nuclear Safety and Security ABET EC2000 syllabus

    E-Print Network [OSTI]

    Ben-Yakar, Adela

    ME 379M- Nuclear Safety and Security Page 1 ABET EC2000 syllabus ME 379M Nuclear Safety assessment models and nuclear non-proliferation. Failure classifications, failure modes, effects, and criticality analysis (FMECA), fault and event trees, reliability block diagrams. Specific areas from the code

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

    SciTech Connect (OSTI)

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

    2011-08-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Koponen, B.L.

    1993-09-20T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Marshall, A.C.; Sawyer, J.C. Jr.; Bari, R.A.; Brown, N.W.; Cullingford, H.S.; Hardy, A.C.; Lee, J.H.; Mcculloch, W.H.; Niederauer, G.F.; Remp, K. (Sandia National Laboratories, Albuquerque, NM (United States) NASA, Washington (United States) Brookhaven National Laboratory, Upton, NY (United States) General Electric Co., San Jose, CA (United States) NASA, Johnson Space Center, Houston, Tn (United States) L

    1992-07-01T23:59:59.000Z

    An interagency Nuclear Safety Policy Working Group (NSPWG) was chartered to recommend nuclear safety policy, requirements, and guidelines for the Space Exploration Initiative (SEI) nuclear propulsion program to facilitate the implementation of mission planning and conceptual design studies. The NSPWG developed a top-level policy to provide the guiding principles for the development and implementation of the nuclear propulsion safety program and the development of safety functional requirements. In addition, the NSPWG reviewed safety issues for nuclear propulsion and recommended top-level safety requirements and guidelines to address these issues. Safety requirements were developed for reactor start-up, inadvertent criticality, radiological release and exposure, disposal, entry, and safeguards. Guidelines were recommended for risk/reliability, operational safety, flight trajectory and mission abort, space debris and meteoroids, and ground test safety. In this paper the specific requirements and guidelines will be discussed. 9 refs.

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

    SciTech Connect (OSTI)

    Marshall, A.C.; Lee, J.H.; McCulloch, W.H. (Sandia National Labs., Albuquerque, NM (United States)); Sawyer, J.C. Jr. (National Aeronautics and Space Administration, Washington, DC (United States)); Bari, R.A. (Brookhaven National Lab., Upton, NY (United States)); Brown, N.W. (General Electric Co., San Jose, CA (United States)); Cullingford, H.S.; Hardy, A.C. (National Aeronautics and Space Administ

    1992-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1992-09-01T23:59:59.000Z

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

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

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

    Nuclear Safety Research and Development Program Proposal Submittal Instructions for Fiscal Year 2015 1.0 INTRODUCTION The Nuclear Safety Research and Development (NSR&D) Program...

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

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

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

  3. Nuclear Explosive Safety - DOE Directives, Delegations, and Requiremen...

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

    2D Admin Chg 1, Nuclear Explosive Safety by Carl Sykes This Department of Energy (DOE) Order establishes requirements to implement the nuclear explosive safety (NES) elements of...

  4. Nuclear Explosive Safety - DOE Directives, Delegations, and Requiremen...

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

    52.2E, Nuclear Explosive Safety by Angela Chambers Functional areas: Safety, Security This Department of Energy (DOE) Order establishes requirements to implement the nuclear...

  5. 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

    2009-09-01T23:59:59.000Z

    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.

  6. Surveillance Guide - NSS 18.1 Criticality Safety

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

    Date: Facility Representative RL Facility Representative Program May 26, 1995 Surveillance Guide NSS-18.1 Revision 0 Criticality Safety Page 5 of 7...

  7. Western University Nuclear Radiation Safety Inspection Checklist

    E-Print Network [OSTI]

    Sinnamon, Gordon J.

    May 2012 Western University Nuclear Radiation Safety Inspection Checklist Permit Holder to nuclear substances or radiation devices is restricted to authorized radiation users listed on the permit radiation labs whenever unsealed nuclear substances are used in these designated locations. 1.7(d

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

    E-Print Network [OSTI]

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

  9. An effective technique for the software requirements analysis of NPP safety-critical systems, based on software inspection, requirements

    E-Print Network [OSTI]

    &V) is therefore emphasized for nuclear safety. Inspection is widely believed to be an effective techniqueAn effective technique for the software requirements analysis of NPP safety-critical systems, based Seonga , Junbeom Yoob , Sung Deok Chab , Yeong Jae Yooc,1 a Department of Nuclear and Quantum Engineering

  10. The Activities of the International Criticality Safety Benchmark Evaluation Project (ICSBEP)

    SciTech Connect (OSTI)

    Briggs, Joseph Blair

    2001-10-01T23:59:59.000Z

    The International Criticality Safety Benchmark Evaluation Project (ICSBEP) was initiated in 1992 by the United States Department of Energy. The ICSBEP became an official activity of the Organization for Economic Cooperation and Development (OECD) – Nuclear Energy Agency (NEA) in 1995. Representatives from the United States, United Kingdom, France, Japan, the Russian Federation, Hungary, Republic of Korea, Slovenia, Yugoslavia, Kazakhstan, Spain, and Israel are now participating. The purpose of the ICSBEP is to identify, evaluate, verify, and formally document a comprehensive and internationally peer-reviewed set of criticality safety benchmark data. The work of the ICSBEP is published as an OECD handbook entitled “International Handbook of Evaluated Criticality Safety Benchmark Experiments”. The 2001 Edition of the Handbook contains benchmark specifications for 2642 critical or subcritical configurations that are intended for use in validation efforts and for testing basic nuclear data.

  11. Neutron absorbing coating for nuclear criticality control

    DOE Patents [OSTI]

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

    2007-10-23T23:59:59.000Z

    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.

  12. COG - Special Features of Interest to Criticality Safety Practitioners

    SciTech Connect (OSTI)

    Buck, R M; Heinrichs, D P; Krass, A W; Lent, E M

    2010-01-14T23:59:59.000Z

    COG is a modern, general-purpose, high fidelity, multi-particle transport code developed at the Lawrence Livermore National Laboratory specifically for use in deep penetration (shielding) and criticality safety calculations. This paper describes some features in COG of special interest to criticality safety practitioners.

  13. Nuclear Explosive Safety Study Functional Area Qualification Standard

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

    2010-05-27T23:59:59.000Z

    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.

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

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

    Change, Safety, The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety,...

  15. 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.

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

    SciTech Connect (OSTI)

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

    2012-01-01T23:59:59.000Z

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

  17. 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.

  18. 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.

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

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

    Implementation, May 31, 2013 (HSS CRAD 45-18, Rev. 1) More Documents & Publications CRAD, Criticality Safety Controls Implementation - May 31, 2013 DOE-STD-1158-2010 Application of...

  20. Towards verifiable adaptive control for safety critical applications

    E-Print Network [OSTI]

    Schwager, Mac

    2005-01-01T23:59:59.000Z

    To be implementable in safety critical applications, adaptive controllers must be shown to behave strictly according to predetermined specifications. This thesis presents two tools for verifying specifications relevant to ...

  1. Influence of safeguards and fire protection on criticality safety

    SciTech Connect (OSTI)

    Six, D E

    1980-01-01T23:59:59.000Z

    There are several positive influences of safeguards and fire protection on criticality safety. Experts in each discipline must be aware of regulations and requirements of the others and work together to ensure a fault-tree design. EG and G Idaho, Inc., routinely uses an Occupancy-Use Readiness Manual to consider all aspects of criticality safety, fire protection, and safeguards. The use of the analytical tree is described.

  2. 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

  3. Java for Safety-Critical Applications Thomas Henties1

    E-Print Network [OSTI]

    Vitek, Jan

    Java for Safety-Critical Applications Thomas Henties1 Siemens AG James J. Hunt aicas Doug Locke Locke Consulting, LLC Kelvin Nilsen Aonix NA Martin Schoeberl Institute of Computer Engineering Vienna for Java in real-time applications with safety requirements has led to an effort to define a new standard

  4. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect (OSTI)

    Sharp, G.L.; McCracken, R.T.

    2003-05-13T23:59:59.000Z

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  5. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect (OSTI)

    Gregg L. Sharp; R. T. McCracken

    2003-06-01T23:59:59.000Z

    The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  6. Determination of Critical Exponents in Nuclear Systems

    E-Print Network [OSTI]

    W. F. J. Mueller; ALADIN collaboration

    1996-07-08T23:59:59.000Z

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

  7. Critical Nuclear Charges for N-Electron Atoms

    E-Print Network [OSTI]

    Kais, Sabre

    Critical Nuclear Charges for N-Electron Atoms ALEXEI V. SERGEEV, SABRE KAIS Department of Chemistry, which is treated as a continuous parameter, approaches its critical value. The critical nuclear charge: critical nuclear charges; N-electron atoms; stability of atomic dianions Introduction he question

  8. Management of National Nuclear Power Programs for assured safety

    SciTech Connect (OSTI)

    Connolly, T.J. (ed.)

    1985-01-01T23:59:59.000Z

    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).

  9. Pantex receives two safety awards | National Nuclear Security...

    National Nuclear Security Administration (NNSA)

    two safety awards | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear...

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

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

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

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

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

    December 8, 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...

  12. Nuclear Facility Safety Basis Fundamentals Self-Study Guide ...

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

    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) November 2002 Nuclear...

  13. Central Technical Authority Responsibilities Regarding Nuclear Safety Requirements

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

    2007-08-28T23:59:59.000Z

    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. No cancellations.

  14. Safety of Decommissioning of Nuclear Facilities

    SciTech Connect (OSTI)

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

    2008-01-15T23:59:59.000Z

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

  15. An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses--Criticality (keff) Predictions

    SciTech Connect (OSTI)

    Scaglione, John M [ORNL] [ORNL; Mueller, Don [ORNL] [ORNL; Wagner, John C [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    One of the most significant remaining challenges associated with expanded implementation of burnup credit in the United States is the validation of depletion and criticality calculations used in the safety evaluation - in particular, the availability and use of applicable measured data to support validation, especially for fission products. Applicants and regulatory reviewers have been constrained by both a scarcity of data and a lack of clear technical basis or approach for use of the data. U.S. Nuclear Regulatory Commission (NRC) staff have noted that the rationale for restricting their Interim Staff Guidance on burnup credit (ISG-8) to actinide-only is based largely on the lack of clear, definitive experiments that can be used to estimate the bias and uncertainty for computational analyses associated with using burnup credit. To address the issue of validation, the NRC initiated a project with the Oak Ridge National Laboratory to (1) develop and establish a technically sound validation approach (both depletion and criticality) for commercial spent nuclear fuel (SNF) criticality safety evaluations based on best-available data and methods and (2) apply the approach for representative SNF storage and transport configurations/conditions to demonstrate its usage and applicability, as well as to provide reference bias results. The purpose of this paper is to describe the criticality (k{sub eff}) validation approach, and resulting observations and recommendations. Validation of the isotopic composition (depletion) calculations is addressed in a companion paper at this conference. For criticality validation, the approach is to utilize (1) available laboratory critical experiment (LCE) data from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and the French Haut Taux de Combustion (HTC) program to support validation of the principal actinides and (2) calculated sensitivities, nuclear data uncertainties, and the limited available fission product LCE data to predict and verify individual biases for relevant minor actinides and fission products. This paper (1) provides a detailed description of the approach and its technical bases, (2) describes the application of the approach for representative pressurized water reactor and boiling water reactor safety analysis models to demonstrate its usage and applicability, (3) provides reference bias results based on the prerelease SCALE 6.1 code package and ENDF/B-VII nuclear cross-section data, and (4) provides recommendations for application of the results and methods to other code and data packages.

  16. 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)]

    2000-03-28T23:59:59.000Z

    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.

  17. Nuclear Explosive Safety Study Process

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin of Contamination in ManyDepartment of Energy NorthB O|WorkNationalNuclear Energyi

  18. Nuclear Safety | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreaking ofOilNEWResponse to Time-Based Rates from theLiabilityEnergyNuclear

  19. Verification of MCNP5-1.60 and MCNP6-Beta-2 for Criticality Safety Applications

    SciTech Connect (OSTI)

    Brown, Forrest B. [Los Alamos National Laboratory; Kiedrowski, Brian C. [Los Alamos National Laboratory; Bull, Jeffrey S. [Los Alamos National Laboratory

    2012-05-01T23:59:59.000Z

    To verify that both MCNP5-1.60 and MCNP6-Beta-2 are performing correctly for criticality safety applications, several suites of verification/validation benchmark problems were run in early 2012. Results from these benchmark suites were compared with results from previously verified versions of MCNP5. The goals of this verification testing were: (1) Verify that MCNP5-1.60 works correctly for nuclear criticality safety applications, producing the same results as for the previous verification performed in 2010; (2) Determine the sensitivity to computer roundoff using different Fortran-90 compilers for building MCNP5 and MCNP6, to support moving to current versions of the compilers; and (3) Verify that MCNP6-Beta-2 works correctly for nuclear criticality safety applications, producing the same results as for MCNP5-1.60. This provides support for eventual migration of users and applications to MCNP6. The current production version of MCNP5 included in the RSICC release package is MCNP5-1.60. This version was first distributed by RSICC in October 2010. While there were subsequent RSICC distributions of the MCNP package in July 2011 and February 2012, no changes were made to MCNP5-1.60. The RSICC release package in February 2012 included both MCNP5-1.60 and the current beta version of MCNP6, MCNP6-Beta-2. MCNP6 is the merger of MCNP5 and MCNPX capabilities. The current release of MCNP6 available from RSICC as of February 2012 is MCNP6-Beta-2. This version includes all of the features for criticality safety calculations that are available in MCNP5-1.60, and many new features largely unrelated to nuclear criticality safety calculations. This release is a 'beta' release to allow intermediate and advanced users to begin testing the merged code in their field of expertise. It should not be used for production calculations.

  20. Computational Methods for Sensitivity and Uncertainty Analysis in Criticality Safety

    SciTech Connect (OSTI)

    Broadhead, B.L.; Childs, R.L.; Rearden, B.T.

    1999-09-20T23:59:59.000Z

    Interest in the sensitivity methods that were developed and widely used in the 1970s (the FORSS methodology at ORNL among others) has increased recently as a result of potential use in the area of criticality safety data validation procedures to define computational bias, uncertainties and area(s) of applicability. Functional forms of the resulting sensitivity coefficients can be used as formal parameters in the determination of applicability of benchmark experiments to their corresponding industrial application areas. In order for these techniques to be generally useful to the criticality safety practitioner, the procedures governing their use had to be updated and simplified. This paper will describe the resulting sensitivity analysis tools that have been generated for potential use by the criticality safety community.

  1. TECHNICAL BASIS FOR THE NUCLEAR CRITICALITY REPRESENTATIVE ACCIDENT & ASSOCIATED REPRESENTED HAZARDOUS CONDITIONS

    SciTech Connect (OSTI)

    GRIGSBY, J.M.

    2005-03-03T23:59:59.000Z

    Technical Basis Document for the Nuclear Criticality Representative Accident and Associate Represented Hazardous Conditions. Revision 2 of RPP-12371 provides accident consequence estimates for a hypothetical criticality event in an above grade facility (e.g. DBVS, CH-TRUM, and S-109 PWRS). This technical basis document was developed to support RPP-13033, ''Tank Farms Documented Safety Analysis (DSA)'', and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls.

  2. Nuclear Criticality Safety | More Science | ORNL

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy:Nanowire3627 Federal Register /7 This is a

  3. Office of Nuclear Safety | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122Commercial602 1,39732onMake YourDepartment ofC TEnergy Nuclear Safety andNuclear

  4. 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).

  5. Double-clad nuclear fuel safety rod

    DOE Patents [OSTI]

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

    1984-01-01T23:59:59.000Z

    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.

  6. 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

  7. Fast Reactor Spent Fuel Processing: Experience and Criticality Safety

    SciTech Connect (OSTI)

    Chad Pope

    2007-05-01T23:59:59.000Z

    This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day operations as well as obtaining historical information. Over 12,000 driver fuel elements have been processed resulting in the production of 2500 kg of 20% enriched uranium. Also, over one thousand blanket fuel elements have been processed resulting in the production of 2400 kg of depleted uranium. These operations required over 35,000 fissile material transfers between zones and over 6000 transfers between containers. Throughout all of these movements, no mass limit violations occurred. Numerous lessons were learned over the ten year operating history. From a criticality safety perspective, the most important lesson learned was the involvement of a criticality safety practitioner in daily operations. A criticality safety engineer was assigned directly to facility operations, and was responsible for implementation of limits and controls including upkeep of the associated computerized tracking files. The criticality safety engineer was also responsible for conducting fuel handler training activities including serving on fuel handler qualification oral boards, and continually assessing operations from a criticality control perspective. The criticality safety engineer also attended bimonthly project planning meetings to identify upcoming process changes that would require criticality safety evaluation. Finally, the excellent criticality safety record was due in no small part to the continual support, involvement, trust, and confidence of project and operations mana

  8. Technical basis for the nuclear criticality representative accident and associated represented hazardous conditions

    SciTech Connect (OSTI)

    CARSON, D.M.

    2003-03-20T23:59:59.000Z

    This technical basis document was developed to support the Tank Farms Documented Safety Analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence. Note that the risk binning process is not applied to facility workers, because all facility worker hazardous conditions are considered for safety-significant SSCs and/or TSR-level controls. Determination of the need for safety-class SSCs was performed in accordance with DOE-STD-3009-94, ''Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', as described in this report.

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

    SciTech Connect (OSTI)

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

    2011-09-01T23:59:59.000Z

    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.

  10. Facility Safety

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

    2000-11-20T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2014-01-01T23:59:59.000Z

    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.

  12. Ris-R-1318(EN) Nuclear Safety Research

    E-Print Network [OSTI]

    . Risø Decommissioning has the task of preparing for the decommissioning of Risø's nuclear facilitiesRisø-R-1318(EN) Nuclear Safety Research Department Annual Report 2001 Edited by B. Majborn, A This report presents a summary of the work of the Nuclear Safety Research Department in 2001. The department

  13. Ris-R-1019(EN) Nuclear Safety Research

    E-Print Network [OSTI]

    .2 Severe accidents 7 2.3 Decommissioning of research reactors 9 2.4 Nuclear information 10 3 RadiationRisø-R-1019(EN) Nuclear Safety Research and Facilities Department Annual Report 1997 Edited by B of the work of the Nuclear Safety Research and Facilities Department in 1997. The department´s research

  14. Ris-R-1104(EN) Nuclear Safety Research

    E-Print Network [OSTI]

    .4 Radioanalytical chemistry 36 3.5 Ecophysiology 38 3.6 Radioactive waste 41 4 Nuclear facilities and services 43 4Risř-R-1104(EN) Nuclear Safety Research and Facilities Department Annual Report 1998 Edited by B, Roskilde, Denmark April 1999 #12;Abstract The report presents a summary of the work of the Nuclear Safety

  15. Ris-M-2944(EN) Nuclear Safety Research

    E-Print Network [OSTI]

    for Maintaining Nuclear Knowledge 23 3.3. DR 1 24 4. THE WASTE MANAGEMENT SECTION 26 4.1. Waste Management 26 4Risř-M-2944(EN) 3> Nuclear Safety Research Department Annual Progress Report 1990 Edited by F. Heikel Vinther Risř National Laboratory, Roskilde, Denmark July 1991 #12;Nuclear Safety Research

  16. Ris-R-1162(EN) Nuclear Safety Research

    E-Print Network [OSTI]

    .4 Radioanalytical chemistry 26 3.5 Radioactive waste 29 4 Nuclear facilities and services 30 4.1 Research reactor DRRisř-R-1162(EN) Nuclear Safety Research and Facilities Department Annual Report 1999 Edited by B, Denmark April 2000 #12;Abstract The report presents a summary of the work of the Nuclear Safety Research

  17. TECHNICAL BASIS FOR THE NUCLEAR CRITICALITY REPRESENTATIVE ACCIDENT & ASSOCIATED REPRESENTED HAZARDOUS CONDITIONS

    SciTech Connect (OSTI)

    GOETZ, T.G.

    2003-06-17T23:59:59.000Z

    This document was developed to support the documented safety analysis (DSA) and describes the process and basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. Revision 1 incorporates ORP IRT comments to enhance the technical presentation and also makes editorial changes. This technical basis document was developed to support the documented safety analysis (DSA), and describes the risk binning process and the technical basis for assigning risk bins for the nuclear criticality representative accident and associated hazardous conditions. The purpose of the risk binning process is to determine the need for safety-significant structures, systems, and components (SSC) and technical safety requirement (TSR)-level controls for a given representative accident or represented hazardous condition based on an evaluation of the frequency and consequence.

  18. Safety program considerations for space nuclear reactor systems

    SciTech Connect (OSTI)

    Cropp, L.O.

    1984-08-01T23:59:59.000Z

    This report discusses the necessity for in-depth safety program planning for space nuclear reactor systems. The objectives of the safety program and a proposed task structure is presented for meeting those objectives. A proposed working relationship between the design and independent safety groups is suggested. Examples of safety-related design philosophies are given.

  19. Implications of Monte Carlo Statistical Errors in Criticality Safety Assessments

    SciTech Connect (OSTI)

    Pevey, Ronald E.

    2005-09-15T23:59:59.000Z

    Most criticality safety calculations are performed using Monte Carlo techniques because of Monte Carlo's ability to handle complex three-dimensional geometries. For Monte Carlo calculations, the more histories sampled, the lower the standard deviation of the resulting estimates. The common intuition is, therefore, that the more histories, the better; as a result, analysts tend to run Monte Carlo analyses as long as possible (or at least to a minimum acceptable uncertainty). For Monte Carlo criticality safety analyses, however, the optimization situation is complicated by the fact that procedures usually require that an extra margin of safety be added because of the statistical uncertainty of the Monte Carlo calculations. This additional safety margin affects the impact of the choice of the calculational standard deviation, both on production and on safety. This paper shows that, under the assumptions of normally distributed benchmarking calculational errors and exact compliance with the upper subcritical limit (USL), the standard deviation that optimizes production is zero, but there is a non-zero value of the calculational standard deviation that minimizes the risk of inadvertently labeling a supercritical configuration as subcritical. Furthermore, this value is shown to be a simple function of the typical benchmarking step outcomes--the bias, the standard deviation of the bias, the upper subcritical limit, and the number of standard deviations added to calculated k-effectives before comparison to the USL.

  20. Nuclear Safety Policy - DOE Directives, Delegations, and Requirements

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

    ARCHIVED SEN-35-91, Nuclear Safety Policy by jnichols Functional areas: Environment, Safety, and Health, Canceled by DOE P 420.1 n3591.pdf -- PDF Document, 20 KB Writer: jnichols...

  1. Critical phenomena of asymmetric nuclear matter in the extended

    E-Print Network [OSTI]

    Critical phenomena of asymmetric nuclear matter in the extended Zimanyi-Moszkowski model K nuclear matter produced by heavy-ion reactions is isospin asymmetric. Although the critical exponents. Miyazaki Abstract We have studied the liquid-gas phase transition of warm asymmetric nuclear matter

  2. Sandia Energy - Sandia Teaches Nuclear Safety Course

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of ScienceandMesa del Sol Home Distribution GridDocuments HomeDatabase on EngineATeaches Nuclear Safety

  3. Nuclear Safety Regulatory Framework | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33Frequently Asked QuestionsDepartment ofDepartment640 FederalDepartment ofNRC'sNuclear Safety

  4. 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

  5. CRITICALITY HAZOP EFFICIENTLY EVALUATING HAZARDS OF NEW OR REVISED CRITICALITY SAFETY EVALUATIONS

    SciTech Connect (OSTI)

    CARSON DM

    2008-04-15T23:59:59.000Z

    The 'Criticality HazOp' technique, as developed at Hanford's Plutonium Finishing Plant (PFP), has allowed for efficiencies enabling shortening of the time necessary to complete new or revised criticality safety evaluation reports (CSERs). For example, in the last half of 2007 at PFP, CSER revisions undergoing the 'Criticality HazOp' process were completed at a higher rate than previously achievable. The efficiencies gained through use of the 'Criticality HazOp' process come from the preliminary narrowing of potential scenarios for the Criticality analyst to fully evaluate in preparation of the new or revised CSER, and from the use of a systematized 'Criticality HazOp' group assessment of the relevant conditions to show which few parameter/condition/deviation combinations actually require analytical effort. The 'Criticality HazOp' has not only provided efficiencies of time, but has brought to criticality safety evaluation revisions the benefits of a structured hazard evaluation method and the enhanced insight that may be gained from direct involvement of a team in the process. In addition, involved personnel have gained a higher degree of confidence and understanding of the resulting CSER product.

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

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

    2014-12-19T23:59:59.000Z

    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.

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

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

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

    Office of Environmental Management (EM)

    examine implementation of safety management programs during phases of the life cycle of a nuclear facility including operation, deactivation, decontamination, decommissioning and...

  9. Nuclear Safety Enforcement Letter issued to Sandia Corporation...

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

    Enforcement Letter issued to Sandia Corporation Nuclear Safety Enforcement Letter issued to Sandia Corporation 9222014 Enforcement Letter, Sandia Corporation, September 22, 2014...

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

    Energy Savers [EERE]

    Safety Research document "MANAGEMENT OF CHANGE IN THE NUCLEAR INDUSTRY - Evidence from maintenance reorganizations," NKS 119, March 2006. xiv | RECOMMENDATIONS Enforcement and...

  11. Spent Nuclear Fuel project integrated safety management plan

    SciTech Connect (OSTI)

    Daschke, K.D.

    1996-09-17T23:59:59.000Z

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

  12. 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.

  13. Application of Covariance Data to Criticality Safety Data Validation

    SciTech Connect (OSTI)

    Broadhead, B.L.; Hopper, C.M.; Parks, C.V.

    1999-11-13T23:59:59.000Z

    The use of cross-section covariance data has long been a key part of traditional sensitivity and uncertainty analyses (S/U). This paper presents the application of S/U methodologies to the data validation tasks of a criticality safety computational study. The S/U methods presented are designed to provide a formal means of establishing the area (or range) of applicability for criticality safety data validation studies. The goal of this work is to develop parameters that can be used to formally determine the ''similarity'' of a benchmark experiment (or a set of benchmark experiments individually) and the application area that is to be validated. These parameters are termed D parameters, which represent the differences by energy group of S/U-generated sensitivity profiles, and ck parameters, which are the correlation coefficients, each of which gives information relative to the similarity between pairs of selected systems. The application of a Generalized Linear Least-Squares Methodology ( GLLSM) tool to criticality safety validation tasks is also described in this paper. These methods and guidelines are also applied to a sample validation for uranium systems with enrichments greater than 5 wt %.

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

    SciTech Connect (OSTI)

    Putman, V.L.

    1995-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2009-02-01T23:59:59.000Z

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

  16. Three Quantities for Error Evaluation in Safety Critical Human Computer Interface 1

    E-Print Network [OSTI]

    Schreiber, Fabio A.

    1 Three Quantities for Error Evaluation in Safety Critical Human Computer Interface 1 Fabio A on the total dependability figures in safety critical systems: usability ­ expressed as a function interface, learning difficulty, MTBF, safety critical systems, usability 1. Introduction The embedding

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

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    CTMCONTROL: Addressing the MC/DC Objective for Safety-Critical Automotive Software Anila Mjeda.mjeda,mike.hinchey}@lero.ie Abstract. We propose a method tailored to the requirements of safety-critical embedded automotive software/DC) objective for automotive safety-critical software. CTMCONTROL is validated via a controlled experiment which

  18. 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.

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

    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.

  20. Ris-R-739(p*tf Nuclear Safety Research

    E-Print Network [OSTI]

    D. i3 Risř-R-739(p*tf Nuclear Safety Research Department Annual Progress Report 1993 Edited by B 1994 #12;Nuclear Safety Research ****-mp. The activities cover health physics, reactor physics, operation of the small reactor DR.1, and radioactive waste

  1. Nuclear Plant/Hydrogen Plant Safety: Issues and Approaches

    SciTech Connect (OSTI)

    Steven R. Sherman

    2007-06-01T23:59:59.000Z

    The U.S. Department of Energy, through its agents the Next Generation Nuclear Plant Project and the Nuclear Hydrogen Initiative, is working on developing the technologies to enable the large scale production of hydrogen using nuclear power. A very important consideration in the design of a co-located and connected nuclear plant/hydrogen plant facility is safety. This study provides an overview of the safety issues associated with a combined plant and discusses approaches for categorizing, quantifying, and addressing the safety risks.

  2. Accurate fission data for nuclear safety

    E-Print Network [OSTI]

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

    2013-04-08T23:59:59.000Z

    The Accurate fission data for nuclear safety (AlFONS) project aims at high precision measurements of fission yields, using the renewed IGISOL mass separator facility in combination with a new high current light ion cyclotron at the University of Jyvaskyla. The 30 MeV proton beam will be used to create fast and thermal neutron spectra for the study of neutron induced fission yields. Thanks to a series of mass separating elements, culminating with the JYFLTRAP Penning trap, it is possible to achieve a mass resolving power in the order of a few hundred thousands. In this paper we present the experimental setup and the design of a neutron converter target for IGISOL. The goal is to have a flexible design. For studies of exotic nuclei far from stability a high neutron flux (10^12 neutrons/s) at energies 1 - 30 MeV is desired while for reactor applications neutron spectra that resembles those of thermal and fast nuclear reactors are preferred. It is also desirable to be able to produce (semi-)monoenergetic neutrons for benchmarking and to study the energy dependence of fission yields. The scientific program is extensive and is planed to start in 2013 with a measurement of isomeric yield ratios of proton induced fission in uranium. This will be followed by studies of independent yields of thermal and fast neutron induced fission of various actinides.

  3. Accurate fission data for nuclear safety

    E-Print Network [OSTI]

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

    2013-01-01T23:59:59.000Z

    The Accurate fission data for nuclear safety (AlFONS) project aims at high precision measurements of fission yields, using the renewed IGISOL mass separator facility in combination with a new high current light ion cyclotron at the University of Jyvaskyla. The 30 MeV proton beam will be used to create fast and thermal neutron spectra for the study of neutron induced fission yields. Thanks to a series of mass separating elements, culminating with the JYFLTRAP Penning trap, it is possible to achieve a mass resolving power in the order of a few hundred thousands. In this paper we present the experimental setup and the design of a neutron converter target for IGISOL. The goal is to have a flexible design. For studies of exotic nuclei far from stability a high neutron flux (10^12 neutrons/s) at energies 1 - 30 MeV is desired while for reactor applications neutron spectra that resembles those of thermal and fast nuclear reactors are preferred. It is also desirable to be able to produce (semi-)monoenergetic neutrons...

  4. FAQS Job Task Analyses - Criticality Safety | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd of Year 2010Salt |ExelonFAQ: RelocationCriticality Safety FAQS Job Task

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

    Office of Environmental Management (EM)

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

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    passive safety cooling systems. To develop an understandingthe passive safety cooling system and recommend an approachof Passive Safety Cooling Systems for Advanced Nuclear

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

    E-Print Network [OSTI]

    Ostroff, Jonathan S.

    Formal Methods for the Specification and Design of Real­Time Safety Critical Systems \\Lambda Jonathan S. Ostroff April, 1992 Abstract Safety critical computers increasingly affect nearly every aspect, designing and verifying real­time sys­ tems, so as to improve their safety and reliability. \\Lambda

  8. Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ-II Safety Program

    SciTech Connect (OSTI)

    Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Garin, V.P.; Gomin, E.A.; Kompanietz, G.V.; Krutoy, A.M.; Lobynstev, V.A.; Maiorov, L.V.; Polyakov, D.N. [RRC Kurchatov Institute, Moscow (Russian Federation)] [and others

    1994-11-01T23:59:59.000Z

    Presented is a brief description of the Narciss-M2 critical assemblies, which simulate accidental water/wet-sand immersion of the TOPAZ-II reactor as well as water-flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel-out anticriticality-device modification to the TOPAZ-II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron-transport method is outlined. Nuclear criticality safety of the TOPAZ-II reactor with an incorporated anticriticality unit is demonstrated.

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

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

    Health, Safety and Security (HSS) observed the public hearing of the DNFSB review of the UPF project status for integrating safety into design. The meeting was broken into three...

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

    Energy Savers [EERE]

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

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

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

    Technical Safety Requirements - December 2, 2014 (EA CRAD 31-07, Rev. 0) More Documents & Publications CRAD, Nuclear Reactor Facility Operations - December 4, 2014 (EA CRAD 31-08...

  12. Preparation Of Nonreactor Nuclear Facility Documented Safety Analysis

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

    2014-11-12T23:59:59.000Z

    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.

  13. Interface with the Defense Nuclear Facilities Safety Board

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

    1996-12-30T23:59:59.000Z

    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.

  14. Interface with the Defense Nuclear Facilities Safety Board

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

    1999-01-26T23:59:59.000Z

    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.

  15. Interface with the Defense Nuclear Facilities Safety Board

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

    2001-03-30T23:59:59.000Z

    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 140.1-1A.

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

    SciTech Connect (OSTI)

    Marshall, A.C.

    1993-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Brewer, Jeffrey D.

    2014-05-01T23:59:59.000Z

    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.

  18. Safety Oversight of Decommissioning Activities at DOE Nuclear Sites

    SciTech Connect (OSTI)

    Zull, Lawrence M.; Yeniscavich, William [Defense Nuclear Facilities Safety Board, 625 Indiana Ave., NW, Suite 700, Washington, DC 20004-2901 (United States)

    2008-01-15T23:59:59.000Z

    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.

  19. The Criticality Safety Information Resource Center at Los Alamos National Laboratory

    SciTech Connect (OSTI)

    Henderson, B.D.; Meade, R.A. [Los Alamos National Lab., NM (United States); Pruvost, N.L. [Galaxy Computer Services, Inc., Santa Fe, NM (United States)

    1997-05-01T23:59:59.000Z

    The mission of the Criticality Safety Information Resource Center (CSIRC) at Los Alamos National Laboratory (LANL) is the preservation of primary documentation supporting criticality safety. In many cases, but not all, this primary documentation consists of experimentalists` logbooks. Experience has shown that the logbooks and other primary information are vulnerable to being discarded. Destruction of these logbooks results in a permanent loss to the criticality safety community.

  20. Pre-orbital criticality safety for the NEPSTP mission

    SciTech Connect (OSTI)

    Sapir, J.; Pelowitz, D.; Streetman, J.R. [Los Alamos National Lab., NM (United States); Glushkov, Y.S.; Ponomarev-Stepnoi, N.N.; Kompanietz, G.V.; Lobynstev, V.A. [Kurchatov Inst., Moscow (Russian Federation)

    1993-11-01T23:59:59.000Z

    In December 1991, the Strategic Defense Initiative Organization (SDIO) proposed investigating whether launching a Russian Topaz-II space nuclear power system could be done safely and within budget constraints. Functional safety requirements developed for the US Topaz mission mandated that the reactor remain subcritical when immersed in water. Topaz-II is an epithermal, enriched-uranium-fueled, NaK- (liquid metal alloy with 22% sodium and 78% potassium) cooled, and zirconium hydride-moderated reactor. A radial beryllium reflector containing 12 rotatable control drums surrounds the core. The authors prepared a computer model of the Topaz reactor that explicitly represented all major reactor components. Initial analyses indicated that in several water-immersion scenarios, the reactor would not remain subcritical. After additional calculations, modifications were proposed that would assure subcriticality under such conditions. This paper describes the analyses and the proposed modifications.

  1. aerospace nuclear safety: Topics by E-print Network

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

    aerospace nuclear safety First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 AEROSPACE SAFETY ADVISORY...

  2. 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.

    2000-04-01T23:59:59.000Z

    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.

  3. A critical review of world jet transport safety

    E-Print Network [OSTI]

    Achtmann, Eric D.

    1995-01-01T23:59:59.000Z

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

  4. autoclave nuclear criticality: Topics by E-print Network

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

    autoclave nuclear criticality First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Nuclear Engineering...

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

    Office of Environmental Management (EM)

    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...

  6. Critical frequency in nuclear chiral rotation

    E-Print Network [OSTI]

    P. Olbratowski; J. Dobaczewski; J. Dudek

    2002-11-25T23:59:59.000Z

    Within the cranked Skyrme-Hartree-Fock approach the self-consistent solutions have been obtained for planar and chiral rotational bands in 132La. It turns out that the chiral band cannot exist below some critical rotational frequency which in the present case equals omega=0.6MeV. The appearance of the critical frequency is explained in terms of a simple classical model of two gyroscopes coupled to a triaxial rigid body.

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

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

    2012-12-04T23:59:59.000Z

    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. Cancels DOE G 420.1-1.

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

    SciTech Connect (OSTI)

    Morris, Tommy J. [Los Alamos National Laboratory

    2012-07-05T23:59:59.000Z

    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.

  9. Facility Safety

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

    2005-12-22T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11T23:59:59.000Z

    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.

  11. FAQS Qualification Card – Nuclear Explosive Safety Study

    Broader source: Energy.gov [DOE]

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

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

    E-Print Network [OSTI]

    Schoeberl, Martin

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

  13. A Model for the Human Computer Interface Evaluation in Safety Critical Computer

    E-Print Network [OSTI]

    Schreiber, Fabio A.

    A Model for the Human Computer Interface Evaluation in Safety Critical Computer Applications Fabio of the IEEE International Conference and Workshop: Engineering of Computer­Based Systems March 1998, Jerusalem, Israel #12; 179 A Model for the Human Computer Interface Evaluation in Safety Critical Computer

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

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01T23:59:59.000Z

    Charges Relating to Nuclear Reactor Safety," 1976, availableissues impor tant to nuclear reactor safety. This report wasstudies of overall nuclear reactor safety have been

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

    E-Print Network [OSTI]

    Nero, A.V.

    2010-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Roberts, E.W.; Edson, J.L.; Udy, A.C. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1996-02-01T23:59:59.000Z

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

  17. Facility Safety

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

    1995-10-13T23:59:59.000Z

    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.

  18. Facility Safety

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

    2005-12-22T23:59:59.000Z

    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.

  19. Facility Safety

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

    2012-12-04T23:59:59.000Z

    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.

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power Administration would likeConstitution4 Department of Energy CarlsbadWinterAnyone whoCritical

  1. Pseudo-critical clusterization in nuclear multifragmentation

    E-Print Network [OSTI]

    Diego Gruyer; J. D. Frankland; R. Botet; M. Ploszajczak; E. Bonnet; A. Chbihi; P. Marini

    2013-08-24T23:59:59.000Z

    In this contribution we show that the biggest fragment charge distribution in central collisions of Xe+Sn leading to multifragmentation is an admixture of two asymptotic distributions observed for the lowest and highest bombarding energies. The evolution of the relative weights of the two components with bombarding energy is shown to be analogous to that observed as a function of time for the largest cluster produced in irreversible aggregation for a finite system. We infer that the size distribution of the largest fragment in nuclear multifragmentation is also characteristic of the time scale of the process, which is largely determined by the onset of radial expansion in this energy range.

  2. Nuclear power safety in central and eastern Europe

    SciTech Connect (OSTI)

    Wilson, R. [Harvard Univ., Cambridge, MA (United States)

    1995-01-01T23:59:59.000Z

    The Chernobyl accident showed the weaknesses in the Soviet approach to safety, particularly of nuclear reactors. Until recently, Western governments, scientists, and engineers did not understand how to help their Russian colleagues make a greater society. This article discusses the two main types of Soviet reactors, their safety problems, and the help Westerners are giving to make them safer. 35 refs., 1 fig., 4 tabs.

  3. Criticality Safety of Low-Enriched Uranium and High-Enriched Uranium Fuel Elements in Heavy Water Lattices

    SciTech Connect (OSTI)

    Pesic, Milan P

    2003-10-15T23:59:59.000Z

    The RB reactor was designed as a natural-uranium, heavy water, nonreflected critical assembly in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, in 1958. From 1962 until 2002, numerous critical experiments were carried out with low-enriched uranium and high-enriched uranium fuel elements of tubular shape, known as the Russian TVR-S fuel assembly type, placed in various heavy water square lattices within the RB cylindrical aluminum tank. Some of these well-documented experiments were selected, described, evaluated, and accepted for inclusion in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments', contributing to the preservation of a rather small number of heavy water benchmark critical experiments.

  4. Nuclear Power - Operation, Safety and Environment 

    E-Print Network [OSTI]

    2011-01-01T23:59:59.000Z

    Today's nuclear reactors are safe and highly efficient energy systems that offer electricity and a multitude of co-generation energy products ranging from potable water to heat for industrial applications. At the same time, ...

  5. Department of Energy Nuclear Safety Policy

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

    2011-02-08T23:59:59.000Z

    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. Cancels SEN-35-91.

  6. Gluon condensation and deconfinement critical density in nuclear matter

    E-Print Network [OSTI]

    M. Baldo; P. Castorina; D. Zappala'

    2004-10-07T23:59:59.000Z

    An upper limit to the critical density for the transition to the deconfined phase, at zero temperature, has been evaluated by analyzing the behavior of the gluon condensate in nuclear matter. Due to the non linear baryon density effects, the upper limit to the critical density, \\rho_c turns out about nine times the saturation density, rho_0 for the value of the gluon condensate in vacuum =0.012 GeV^4. For neutron matter \\rho_c \\simeq 8.5 \\rho_0. The dependence of the critical density on the value of the gluon condensate in vacuum is studied.

  7. Nuclear space power safety and facility guidelines study

    SciTech Connect (OSTI)

    Mehlman, W.F.

    1995-09-11T23:59:59.000Z

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

  8. Critical Temperature for the Nuclear Liquid-Gas Phase Transition

    E-Print Network [OSTI]

    V. A. Karnaukhov; H. Oeschler; S. P. Avdeyev; E. V. Duginova; V. K. Rodionov; A. Budzanowski; W. Karcz; O. V. Bochkarev; E. A. Kuzmin; L. V. Chulkov; E. Norbeck; A. S. Botvina

    2003-02-07T23:59:59.000Z

    The charge distribution of the intermediate mass fragments produced in p (8.1 GeV) + Au collisions is analyzed in the framework of the statistical multifragmentation model with the critical temperature for the nuclear liquid-gas phase transition $T_c$ as a free parameter. It is found that $T_c=20\\pm3$ MeV (90% CL).

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

    SciTech Connect (OSTI)

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

    1997-04-09T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    JENSEN, M.A.

    2005-08-19T23:59:59.000Z

    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.

  12. Documented safety analysis for 209-E Facility critical mass laboratory

    SciTech Connect (OSTI)

    DODD, E.

    2003-04-23T23:59:59.000Z

    This documented safety analysis (DSA) addresses the 209-E facility and associated requirements that apply to 200 Area Facility Deactivation (AFD) activities. The DSA was developed in accordance with DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities''.

  13. 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.

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

    SciTech Connect (OSTI)

    Booker, Paul M.; Brown, Lisa M.

    2005-07-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Roozbehani, Mardavij

    2008-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Ouyang, Meng

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

  17. Software reliability and safety in nuclear reactor protection systems

    SciTech Connect (OSTI)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1993-11-01T23:59:59.000Z

    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.

  18. Floating nuclear power plant safety assurance principles

    SciTech Connect (OSTI)

    Zvonarev, B.M.; Kuchin, N.L.; Sergeev, I.V.

    1993-12-31T23:59:59.000Z

    In the north regions of the Russian federation and low density population areas, there is a real necessity for ecological clean energy small power sources. For this purpose, floating nuclear power plants, designed on the basis of atomic ship building engineering, are being conceptualized. It is possible to use the ship building plants for the reactor purposes. Issues such as radioactive waste management are described.

  19. Q)Tf(^/7^,\\ Ris-R-625(pff Nuclear Safety Research

    E-Print Network [OSTI]

    and nuclear releases. fields of radiation protection, reactor safety and radioactive waste managementQ)Tf(^/7^,\\ Risř-R-625(pff Nuclear Safety Research Department Annual Progress Report 1991 Edited Roskilde, Denmark March 1992 #12;Nuclear Safety Research Department Annual Progress Report 1991 Riso-R-62S

  20. Nuclear Safety Reporting Criteria | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin of Contamination in ManyDepartment of Energy NorthBDepartment of Energy Nuclear

  1. Sandia Energy - Nuclear Energy Safety Technologies

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclear PressLaboratory Fellows Jerry

  2. Nuclear safety | Princeton Plasma Physics Lab

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparencyDOE Project TapsDOERecoveryNuclearLife Cycleenergysafety

  3. Water/sand flooded and immersed critical experiment and analysis performed in support of the TOPAZ-II safety program

    SciTech Connect (OSTI)

    Glushkov, E.S.; Ponomarev-Stepnoi, N.N.; Bubelev, V.G.; Garin, V.P.; Gomin, E.A.; Kompanietz, G.V.; Krutov, A.M.; Lobynstev, V.A.; Maiorov, L.V.; Polyakov, D.N.; Chunyaev, E.I. [RRC Kurchatov Institute, Moscow 123182 (Russian Federation); Marshall, A.C. [International Nuclear Safety, Sandia National Laboratories, Albuquerque, New Mexico 87185 (United States); Sapir, J.L.; Pelowitz, D.B. [Reactor Design and Analysis Group, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States)

    1995-01-20T23:59:59.000Z

    Presented is a brief description of the Narciss-M2 critical assemblies, which simulate accidental water/wet-sand immersion of the TOPAZ-II reactor as well as water-flooding of core cavities. Experimental results obtained from these critical assemblies, including experiments with several fuel elements removed from the core, are shown. These configurations with several extracted fuel elements simulate a proposed fuel-out anticriticality-device modification to the TOPAZ-II reactor. Preliminary computational analysis of these experiments using the Monte Carlo neutron-transport method is outlined. Nuclear criticality safety of the TOPAZ-II reactor with an incorporated anticriticality unit is demonstrated. {copyright} {ital 1995} {ital American} {ital Institute} {ital of} {ital Physics}

  4. Understanding Failures in International Safety-Critical Infrastructures: A Comparison of European and North American Power Failures

    E-Print Network [OSTI]

    Johnson, Chris

    of Glasgow, Scotland, UK. Keywords: Critical Infrastructures, Blackouts, Power Distribution, V2 Analysis. Abstract The increasing integration of safety-critical infrastructures introduces vulnerabilities-critical systems creates vulnerabilities across the power distribution, water supply and transportation industries

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

    E-Print Network [OSTI]

    Robert Bruce Hayes

    2013-01-15T23:59:59.000Z

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

  6. 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...

  7. Critical phenomena of nuclear matter in the extended Zimanyi-Moszkowski model

    E-Print Network [OSTI]

    Critical phenomena of nuclear matter in the extended Zimanyi-Moszkowski model K. Miyazaki Abstract in nuclear multifragmentation reactions and the critical temperature has been derived as TC = 20 3 MeV in Ref] to estimate the critical temperature for in...nite nuclear matter, that is, TC = 16:6 0:86 Me

  8. Double-clad nuclear-fuel safety rod

    DOE Patents [OSTI]

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

    1981-12-30T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1993-02-01T23:59:59.000Z

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

  10. FAQS Reference Guide - Criticality Safety | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd of Year 2010Salt |ExelonFAQ:DepartmentDepartment ofEnergySafety

  11. Iterative methods for solving nonlinear problems of nuclear reactor criticality

    SciTech Connect (OSTI)

    Kuz'min, A. M., E-mail: mephi.kam@mail.ru [National Research Nuclear University MEPhI (Russian Federation)

    2012-12-15T23:59:59.000Z

    The paper presents iterative methods for calculating the neutron flux distribution in nonlinear problems of nuclear reactor criticality. Algorithms for solving equations for variations in the neutron flux are considered. Convergence of the iterative processes is studied for two nonlinear problems in which macroscopic interaction cross sections are functionals of the spatial neutron distribution. In the first problem, the neutron flux distribution depends on the water coolant density, and in the second one, it depends on the fuel temperature. Simple relationships connecting the vapor content and the temperature with the neutron flux are used.

  12. Critical review of deeply bound kaonic nuclear states

    E-Print Network [OSTI]

    V. K. Magas; E. Oset; A. Ramos; H. Toki

    2006-11-28T23:59:59.000Z

    We critically revise the recent claims of narrow deeply bound kaonic states and show that at present there is no convincing experimental evidence for their existence. In particular, we discuss in details the claim of K- pp deeply bound state associated to a peak seen in the Lambda p invariant mass spectrum from K- nuclear absorption reactions by the FINUDA collaboration. An explicit theoretical simulation shows that the peak is simply generated from a two-nucleon absorption process, like K- pp --> Lambda p, followed by final-state interactions of the produced particles with the residual nucleus.

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

    SciTech Connect (OSTI)

    Not Available

    1992-08-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Ajo-Franklin, Jonathan

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

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

    E-Print Network [OSTI]

    Heljanko, Keijo

    C4.2 Formal Verification of Safety I&C System Designs: Two Nuclear Power Plant Related Applications and control (I&C) systems play a crucial role in the operation of nuclear power plants (NPP) and other safety is available. The use of model checking to verify two nuclear power plant related systems is described: an arc

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

    E-Print Network [OSTI]

    Chen, Sheng

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

  17. Major safety and operational concerns for fuel debris criticality control

    SciTech Connect (OSTI)

    Tonoike, K.; Sono, H.; Umeda, M.; Yamane, Y.; Kugo, T.; Suyama, K. [Fukushima Project Team, Japan Atomic Energy Agency Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan)

    2013-07-01T23:59:59.000Z

    It can be seen from the criticality control viewpoint that the requirement divides the decommissioning work into two parts. One is the present condition where it is requested to prevent criticality and to monitor subcritical condition while the debris is untouched. The other is future work where the subcritical condition shall be ensured even if the debris condition is changed intentionally by raising water level, debris retrieval, etc. Repair of damages on the containment vessel (CV) walls is one of the most important objectives at present in the site. On completion of this task, it will become possible to raise water levels in the CVs and to shield the extremely high radiation emitted from the debris but there is a dilemma: raising the water level in the CVs implies to bring the debris closer to criticality because of the role of water for slowing down neutrons. This may be solved if the coolant water will start circulating in closed loops, and if a sufficient concentration of soluble neutron poison (borated water for instance) will be introduced in the loop. It should be still noted that this solution has a risk of worsening corrosion of the CV walls. Design of the retrieval operation of debris should be proposed as early as possible, which must include a neutron poison concentration required to ensure that the debris chunk is subcritical. In parallel, the development of the measurement system to monitor subcritical condition of the debris chunk should be conducted in case the borated water cannot be used continuously. The system would be based on a neutron counter with a high sensitivity and an appropriate shield for gamma-rays, and the adequate statistical signal processing.

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

    SciTech Connect (OSTI)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Fisk, Patricia; Rutherford, Lavon

    2003-06-01T23:59:59.000Z

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

  20. Advanced Test Reactor Critical Facility safety analysis report five year currency review

    SciTech Connect (OSTI)

    Napper, P.R.; Carpenter, W.R.; Garner, R.W.

    1991-05-01T23:59:59.000Z

    By DOE-ID Order 5481.1A, a five year currency review is required of the Safety Analysis Reports of all ID or ID contractor operations having hazards of a type and magnitude not routinely encountered and/or accepted by the public. In keeping with this order, a currency review has been performed of the Advanced Test Reactor Critical Facility (ADTRC) Safety Analysis Report (SAR), Issue 003, 1990. The objectives of this currency review were to: evaluate the content, completeness, clarity of presentation and compliance with NRC Regulatory Guides and DOE Orders, etc., and evaluate the technical content of the SAR, particularly the Technical Specifications, and to evaluate the safety of continued operation of the ATRC. The reviewers concluded that although improvements may be needed in the overall content, clarity, and demonstration of compliance with current orders and regulations, the safety of the ATRC is in no way compromised and no unreviewed safety questions were identified. 6 figs., 3 tabs.

  1. Covariance matrices for use in criticality safety predictability studies

    SciTech Connect (OSTI)

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

    1997-09-01T23:59:59.000Z

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

    E-Print Network [OSTI]

    Alexander, Erin L

    2004-01-01T23:59:59.000Z

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

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

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

    to Y-12 in February, 1949. It was a 1,000 KV system installed in Building 9981-1 in the heart of what is known even today as "Area 5." That was but the first of many such X-ray...

  4. Nuclear Criticality Safety Guide for Fire Protection | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin of Contamination in ManyDepartment of Energy NorthB O|Work ForceNovemberof

  5. SURVEYS OF ORGANIZATIONAL CULTURE AND SAFETY CULTURE IN NUCLEAR POWER

    SciTech Connect (OSTI)

    BROWN,B.S.

    2000-07-30T23:59:59.000Z

    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 distinguish among groups within the organization that would be expected to differ.

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd of YearFLASH2011-17-OPAMDepartment ofEnergy Chief of Nuclear Safety

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122Commercial602 1,39732onMake YourDepartment of EnergyNoticeFacilityNuclearSafety

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613 122Commercial602 1,39732onMake YourDepartment ofC TEnergy Nuclear Safety and

  9. Nuclear and Facility Safety Directives | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreaking ofOilNEWResponse to Time-Based Rates fromNuclear Securityand Facility Safety

  10. 2012 Nuclear Safety Workshop Photos | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2UraniumEnergy CityDepartmentDepartment ofDepartment2 DOE2012 Nuclear Safety

  11. Criticality safety concerns of uranium deposits in cascade equipment

    SciTech Connect (OSTI)

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

    1996-12-31T23:59:59.000Z

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

  12. Towards Component-Based Design of Safety-Critical Cyber-Physical Applications

    E-Print Network [OSTI]

    for real-time behavior which is essential in many safety-critical applications. To overcome this problem-physical systems (CPS) can be found in many different domains such as smart traffic and transporta- tion, intelligent buildings, smart grid, etc. A common aspect of such CPS is that they rely on a large number

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

    E-Print Network [OSTI]

    Wedde, Horst F.

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

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

    E-Print Network [OSTI]

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

  15. 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...

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

    SciTech Connect (OSTI)

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

    2006-07-01T23:59:59.000Z

    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)

  17. Underwater nuclear power plants: improved safety, environmental compatibility and efficiency

    SciTech Connect (OSTI)

    Galustov, K.Z.; Abadjyan, K.A.; Pavlov, A.B.

    1991-01-01T23:59:59.000Z

    The further development of nuclear power engineering depends on the creation of a new generation of nuclear power plant (NPP) projects that have a high degree of safety. Decisions ensuring secure NPP exploitation must be based on the possibility of eliminating or localizing accidents. Using environmental properties to achieve secure NPP exploitation and accident elimination leads to suggest the construction of NPPs in water. An efficient way to provide energy to remote coastal areas is through use of floatable construction of prefabricated units. Floatable construction raises the quality of works, reduces expenditures on industrial facilities, and facilities building conditions in districts with extreme climatic conditions. A type of NPP that is situated on a shelf with the reactor compartment placed at the sea bottom is proposed. The underwater location of the reactor compartment on the fixed depth allows the natural water environment conditions of natural hydrostatic pressure, heat transfer and circulation to provide NPP safety. An example of new concept for power units with under-water localization of the reactor compartment is provided by the double-block NPP in a VVER reactor.

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

    E-Print Network [OSTI]

    Crawford, T. Daniel

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

  19. Nuclear Data Capabilities Supported by the DOE NCSP

    E-Print Network [OSTI]

    Danon, Yaron

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

  20. Nonextensive critical effects in relativistic nuclear mean field models

    E-Print Network [OSTI]

    J. Rozynek; G. Wilk

    2011-02-22T23:59:59.000Z

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

  1. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect (OSTI)

    Samuel Bays; Ayodeji Alajo

    2010-05-01T23:59:59.000Z

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

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

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

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

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

    Office of Science (SC) Website

    simulations supporting national security, nonproliferation, advanced reactor and fuel cycle concepts, criticality safety, fusion, space applications, nuclear astrophysics...

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

    SciTech Connect (OSTI)

    none,

    1983-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1991-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    , just as his early work in experiments for nuclear criticality safety for Y-12 impacted national in nuclear criticality safety, nuclear weapons identification, nuclear materials processing, and nuclearJohn T. Mihalczo PhD. in Nuclear Engineering : University of Tennessee, 1970 Masters in Physics

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

    Energy Savers [EERE]

    3116 Public Meeting Summaries - July 2007 Fifth National Report for the Joint Convention on the Safety of Spent Fuel Management and the Safety of Radioactive Waste Management...

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

    SciTech Connect (OSTI)

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

    1993-09-01T23:59:59.000Z

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

  9. Uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. Final report, February 16, 1990--December 31, 1994

    SciTech Connect (OSTI)

    Busch, R.D. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering

    1995-02-24T23:59:59.000Z

    Dr. Robert Busch of the Department of Chemical and Nuclear Engineering was the principal investigator on this project with technical direction provided by the staff in the Nuclear Criticality Safety Group at Los Alamos. During the period of the contract, he had a number of graduate and undergraduate students working on subtasks. The objective of this work was to develop information on uranium systems to enhance benchmarks for use in the verification of criticality safety computer models. During the first year of this project, most of the work was focused on setting up the SUN SPARC-1 Workstation and acquiring the literature which described the critical experiments. By august 1990, the Workstation was operational with the current version of TWODANT loaded on the system. MCNP, version 4 tape was made available from Los Alamos late in 1990. Various documents were acquired which provide the initial descriptions of the critical experiments under consideration as benchmarks. The next four years were spent working on various benchmark projects. A number of publications and presentations were made on this material. These are briefly discussed in this report.

  10. Myths and representations in French nuclear history: The impact on decommissioning safety

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Myths and representations in French nuclear history: The impact on decommissioning safety C. Martin. The decommissioning of many operational plants (whether because nuclear power is being withdrawn, or because plants accident at the Fukushima nuclear power plant has shown that in many countries the debate on the withdrawal

  11. ASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM-OF-SYSTEMS

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe stateASSESSING NUCLEAR POWER PLANT SAFETY AND RECOVERY FROM EARTHQUAKES USING A SYSTEM in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant

  12. Nuclear magnetic ordering in Ca(OH)2. III. Experimental determination of the critical temperature

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1479 Nuclear magnetic ordering in Ca(OH)2. III. Experimental determination of the critical(OH)2 is presented. The ordered phase is reached via dynamic nuclear polarization followed to the effective magnetic field are used to determine the magnetic phase of the nuclear spin system. From

  13. ccsd00001474, Model-independent tracking of criticality signals in nuclear multifragmentation data

    E-Print Network [OSTI]

    ccsd­00001474, version 2 ­ 6 Sep 2004 Model-independent tracking of criticality signals in nuclear Physique Nucléaire, IN2P3-CNRS et Université F-69622 Villeurbanne, France. 8 Institute of Nuclear Physics, Pl-31342 Kraków, Poland. 9 National Institute for Physics and Nuclear Engineering, RO-76900 Bucharest

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

    SciTech Connect (OSTI)

    Rutherford, D.

    1994-03-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

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

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

    SciTech Connect (OSTI)

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

    1982-05-20T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2007-08-01T23:59:59.000Z

    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.

  18. ANS 2006 WINTER MEETING & Nuclear Technology Expo

    E-Print Network [OSTI]

    Krings, Axel W.

    Development Workshop: Digital Instrumentation Upgrades 52 DOE Nuclear Criticality Safety Program 53ANS 2006 WINTER MEETING & Nuclear Technology Expo "Ensuringthe (TOFE) 5th International Topical Meeting on Nuclear Plant Instrumentation, Controls, and Human Machine

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

    SciTech Connect (OSTI)

    Not Available

    1994-10-15T23:59:59.000Z

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

  20. Critical review of the reactor-safety study radiological health effects model. Final report

    SciTech Connect (OSTI)

    Cooper, D.W.; Evans, J.S.; Jacob, N.; Kase, K.R.; Maletskos, C.J.; Robertson, J.B.; Smith, D.G.

    1983-03-01T23:59:59.000Z

    This review of the radiological health effects models originally presented in the Reactor Safety Study (RSS) and currently used by the US Nuclear Regulatory Commission (NRC) was undertaken to assist the NRC in determining whether or not to revise the models and to aid in the revision, if undertaken. The models as presented in the RSS and as implemented in the CRAC (Calculations of Reactor Accident Consequences) Code are described and critiqued. The major elements analyzed are those concerning dosimetry, early effects, and late effects. The published comments on the models are summarized, as are the important findings since the publication of the RSS.

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

    SciTech Connect (OSTI)

    NONE

    1995-12-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1993-09-01T23:59:59.000Z

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

  3. Stability of Relativistic Matter with Magnetic Fields for Nuclear Charges up to the Critical Value

    E-Print Network [OSTI]

    Rupert L. Frank; Elliott H. Lieb; Robert Seiringer

    2006-10-24T23:59:59.000Z

    We give a proof of stability of relativistic matter with magnetic fields all the way up to the critical value of the nuclear charge $Z\\alpha=2/\\pi$.

  4. 1Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory RPI LINAC Facility

    E-Print Network [OSTI]

    Danon, Yaron

    Nuclear Criticality Safety Program Conference April 27, 2011 #12;2Mechanical, Aerospace and Nuclear, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory RPI LINAC Facility Nuclear Criticality Safety1Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory RPI LINAC Facility

  5. Criticality Safety Evaluations on the Use of 200-gram Pu Mass Limit for RHWM Waste Storage Operations

    SciTech Connect (OSTI)

    Chou, P

    2011-12-14T23:59:59.000Z

    This work establishes the criticality safety technical basis to increase the fissile mass limit from 120 grams to 200 grams for Type A 55-gallon drums and their equivalents. Current RHWM fissile mass limit is 120 grams Pu for Type A 55-gallon containers and their equivalent. In order to increase the Type A 55-gallon drum limit to 200 grams, a few additional criticality safety control requirements are needed on moderators, reflectors, and array controls to ensure that the 200-gram Pu drums remain criticality safe with inadvertent criticality remains incredible. The purpose of this work is to analyze the use of 200-gram Pu drum mass limit for waste storage operations in Radioactive and Hazardous Waste Management (RHWM) Facilities. In this evaluation, the criticality safety controls associated with the 200-gram Pu drums are established for the RHWM waste storage operations. With the implementation of these criticality safety controls, the 200-gram Pu waste drum storage operations are demonstrated to be criticality safe and meet the double-contingency-principle requirement per DOE O 420.1.

  6. Critical temperature of antikaon condensation in nuclear matter

    E-Print Network [OSTI]

    Sarmistha Banik; Walter Greiner; Debades Bandyopadhyay

    2008-12-30T23:59:59.000Z

    We investigate the critical temperature of Bose-Einstein condensation of $K^-$ mesons in neutron star matter. This is studied within the framework of relativistic field theoretical models at finite temperature where nucleon-nucleon and (anti)kaon-nucleon interactions are mediated by the exchange of mesons. The melting of the antikaon condensate is studied for different values of antikaon optical potential depths. We find that the critical temperature of antikaon condensation increases with baryon number density. Further it is noted that the critical temperature is lowered as antikaon optical potential becomes less attractive. We also construct the phase diagram of neutron star matter with $K^-$ condensate.

  7. Department of Energy Office of Nuclear Safety and Environmental...

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

    Safety, must comply with national consensus industry standards and the model building codes applicable for the state or region in which the facility is located. Certain...

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

    Office of Environmental Management (EM)

    require component and materials replacement identified and implemented? * Are appropriate preventive maintenance requirements for stored safety-related equipment identified and...

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

    Broader source: Energy.gov [DOE]

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

  10. Safety of nuclear power reactors in the former Eastern European countries

    SciTech Connect (OSTI)

    Chakraborty, S. [Swiss Federal Nuclear Safety Inspectorate, Villigen (Switzerland)

    1995-01-01T23:59:59.000Z

    This article discusses the safety of nuclear power plants in the former Eastern European countries (including the former Soviet Union). The current international design fabrication, construction, operation, safely, regulatory standards and practices, and ways to resolve plant problems are addressed in light of experience with the Western nuclear power development programs. 9 refs., 4 figs.

  11. Fragile Signs of Criticality in the Nuclear Multifragmentation

    E-Print Network [OSTI]

    K. K. Gudima; M. Ploszajczak; V. D. Toneev

    2001-06-08T23:59:59.000Z

    Deviations from an idealized equilibrium phase transition picture in nuclear multifragmentation is studied in terms of the entropic index. We investigate different heat-capacity features in the canonical quantum statistical model of nuclear multifragmentation generalized in the framework of Tsallis nonextensive thermostatistics. We find that the negative branch of heat capacity observed in quasi-peripheral Au+Au collisions is caused primarily by the non-generic nonextensivity effects.

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

    E-Print Network [OSTI]

    2007-01-01T23:59:59.000Z

    improvement in safety based on Institute of Nuclear Power Operations (INPO) performance indicatorsReliability Engineering and System Safety 92 (2007) 609­618 The nuclear industry's transition a Nuclear Science and Engineering Department, Massachusetts Institute of Technology, Cambridge, MA 02139

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

    SciTech Connect (OSTI)

    Fellinger, A.

    2009-10-15T23:59:59.000Z

    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.

  14. Safety of Department of Energy-Owned Nuclear Reactors

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

    1986-09-23T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Ronald E. Mizia

    2011-10-01T23:59:59.000Z

    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.

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

    Broader source: Energy.gov [DOE]

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

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

    SciTech Connect (OSTI)

    VINCENT, Andrew

    2005-07-14T23:59:59.000Z

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

  18. Critical Behavior in Light Nuclear Systems: Experimental Aspects

    E-Print Network [OSTI]

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

    2005-04-14T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1983-08-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    David S. Shafer; Jenna Gommes

    2008-09-15T23:59:59.000Z

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

  1. Aging of Class 1E batteries in safety systems of nuclear power plants

    SciTech Connect (OSTI)

    Edson, J.L.; Hardin, J.E.

    1987-07-01T23:59:59.000Z

    This report presents the results of a study of aging effects on safety-related batteries in nuclear power plants. The purpose is to evaluate the aging effects caused by operation within a nuclear facility and to evaluate maintenance, testing, and monitoring practices with respect to their effectiveness in detecting and mitigating the effects of aging. The study follows the US Nuclear Regulatory Commission's (NRC's) Nuclear Plant-Aging Research approach and investigates the materials used in battery construction, identifies stressors and aging mechanisms, presents operating and testing experience with aging effects, analyzes battery-failure events reported in various data bases, and evaluates recommended maintenance practices. Data bases that were analyzed included the NRC's Licensee Event Report system, the Institute for Nuclear Power Operations' Nuclear Plant Reliability Data System, the Oak Ridge National Laboratory's In-Plant Reliability Data System, and The S.M. Stoller Corporation's Nuclear Power Experience data base.

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

    SciTech Connect (OSTI)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01T23:59:59.000Z

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  3. Microprocessor-based control systems application in nuclear power plant critical systems

    SciTech Connect (OSTI)

    Shah, M.R.; Nowak, J.B. (Sargent and Lundy, Chicago, IL (US))

    1992-01-01T23:59:59.000Z

    Microprocessor-based control systems have been used in fossil power plants and are receiving greater acceptance for application in nuclear plants. This technology is not new but it does require unique considerations when applied to nuclear power plants. Sargent and Lundy (S and L) has used a microprocessor-based component logic control system (interposing Logic System) for safety- and non-safety-related components in nuclear power plants under construction overseas. Currently, S and L is in the design stage to replace an existing analog control system with a microprocessor-based control system in the U.S. The trend in the industry is to replace systems in existing plants or design new power plants with microprocessor-based control systems.

  4. 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.

    1982-06-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    DuPont, John N.

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

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

    SciTech Connect (OSTI)

    Bronisz, S.E. (comp.)

    1983-10-01T23:59:59.000Z

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

  7. The critical temperature of nuclear matter and fragment distributions in multifragmentation of finite nuclei

    E-Print Network [OSTI]

    R. Ogul; A. S. Botvina

    2002-10-15T23:59:59.000Z

    The fragment production in multifragmentation of finite nuclei is affected by the critical temperature of nuclear matter. We show that this temperature can be determined on the basis of the statistical multifragmentation model (SMM) by analyzing the evolution of fragment distributions with the excitation energy. This method can reveal a decrease of the critical temperature that, e.g., is expected for neutron-rich matter. The influence of isospin on fragment distributions is also discussed.

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

    SciTech Connect (OSTI)

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

    2007-10-01T23:59:59.000Z

    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.

  9. Criticality benchmark calculations using PARTISN: Comparisons using MENDF5 and MENDF6 nuclear data libraries.

    SciTech Connect (OSTI)

    Ellis, Ronald J.; Yugo, James J.; Frankle, S. C. (Stephanie C.); Little, R. C. (Robert C.)

    2003-01-01T23:59:59.000Z

    A project was undertaken to assess the MENDF5 and MENDF6 nuclear data libraries through the analysis of 86 critical assembly benchmarks using the LANL discrete ordinates transport code PARTISN. As an initial analysis of the effects of some limitations in the MENDF libraries, this current work assesses differences in k,,a calculations between the PARTISN cases (with MENDF5 and MENDF6 nuclear data libraries) and MCNP cases, and compares these results to the experimental data.

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

    SciTech Connect (OSTI)

    Perkins, W.C.; Durant, W.S.; Dexter, A.H.

    1980-12-01T23:59:59.000Z

    The occurrence of certain potential events in nuclear fuel reprocessing plants could lead to significant consequences involving risk to operating personnel or to the general public. This document is a compilation of such potential initiating events in nuclear fuel reprocessing plants. Possible general incidents and incidents specific to key operations in fuel reprocessing are considered, including possible causes, consequences, and safety features designed to prevent, detect, or mitigate such incidents.

  11. 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.

  12. The impact of offsite factors on the safety performance of small nuclear power plants

    SciTech Connect (OSTI)

    Baranaev, Yu.D.; Viktorov, A.N. [Inst. of Physics and Power Engineering, Obninsk (Russian Federation)

    1991-01-01T23:59:59.000Z

    The results of an analysis of the influence of offsite factors on small nuclear power-plant (SNPP) safety performance during postulated severe accidents are presented. Given the plant locations in the immediate vicinity of residential areas and the impossibility of accomplishing the expeditious evacuation of the public, the risk caused by an SNPP severe accident may be considerably less than that for such an event in a large nuclear power plant. 3 refs., 3 figs., 5 tabs.

  13. Composite neutron absorbing coatings for nuclear criticality control

    DOE Patents [OSTI]

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

    2005-07-19T23:59:59.000Z

    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.

  14. ON THE USE OF SPRAY SYSTEMS: AN EXAMPLE OF R&D WORK IN HYDROGEN SAFETY FOR NUCLEAR APPLICATIONS

    E-Print Network [OSTI]

    Boyer, Edmond

    1 ON THE USE OF SPRAY SYSTEMS: AN EXAMPLE OF R&D WORK IN HYDROGEN SAFETY FOR NUCLEAR APPLICATIONS systems related to hydrogen safety in nuclear power plants have been the subject of several experimental l'Energie Atomique, CEA Saclay, DEN/DM2S/SFME, 91191 Gif-sur-Yvette Cedex, France 2 Institut de

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

    E-Print Network [OSTI]

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

    2008-01-29T23:59:59.000Z

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

  16. CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

    SciTech Connect (OSTI)

    Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

    1982-01-01T23:59:59.000Z

    Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed.

  17. Oak Ridge Critical Experiment Facility (Building 9213)

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

    Joe has continued to hold up this facility as one of the key elements in today's nuclear criticality safety program worldwide. He attributes much of what is routine about...

  18. Y-12 hosts safety workshop | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA Approved: 5-13-14Russian NuclearNational Nuclear Securitysafety workshop |

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

    SciTech Connect (OSTI)

    Forsberg, C.W.; Reich, W.J.

    1991-09-01T23:59:59.000Z

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

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

    E-Print Network [OSTI]

    Maroncelli, Mark

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

  1. International Symposium on Fusion Nuclear Technology (ISFNT-5) SAFETY ISSUES ASSOCIATED WITH MOBILIZED ACTIVATION

    E-Print Network [OSTI]

    California at Los Angeles, University of

    International Symposium on Fusion Nuclear Technology (ISFNT-5) SAFETY ISSUES ASSOCIATED WITH MOBILIZED ACTIVATION PRODUCTS IN SELECTED APEX DESIGNS K. A. McCarthy, D. A. Petti, R. L. Moore, and B. J. In this paper we concentrate on mobilization of first wall materials during ingress events, and provide guidance

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

    SciTech Connect (OSTI)

    Pete Jordan

    2010-09-01T23:59:59.000Z

    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.

  3. Dynamical chaos and critical behavior in Vlasov simulations of nuclear multifragmentation

    E-Print Network [OSTI]

    A. Atalmi; M. Baldo; G. F. Burgio; A. Rapisarda

    1996-02-26T23:59:59.000Z

    We discuss the presence of both dynamical chaos and signals of a second--order phase transition in numerical Vlasov simulations of nuclear multifragmentation. We find that chaoticity and criticality are strongly related and play a crucial role in the process of fragments formation. This connection is not limited to our model and seems a rather general feature.

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

    SciTech Connect (OSTI)

    George, T.G. (comp.)

    1988-07-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    George, T.G. (comp.)

    1989-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Bronisz, S.E. (comp.)

    1983-11-01T23:59:59.000Z

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

  7. Productivity Techniques and Quality Aspects in the Criticality Safety Evaluation of Y-12 Type-B Fissile Material Packages

    SciTech Connect (OSTI)

    DeClue, J. F.

    2011-06-28T23:59:59.000Z

    The inventory of certified Type-B fissile material packages consists of ten performance-based packages for offsite transportation purposes, serving transportation programs at the Y-12 National Security Complex. The containment vessels range from 5 to 19 in. in diameter and from 17 to 58 in. in height. The drum assembly external to the containment vessel ranges from 18 to 34 in. in diameter and from 26 to 71 in. in height. The weight of the packaging (drum assembly and containment vessel) ranges from 239 to 1550 lb. The older DT-nn series of Cellotex-based packages are being phased-out and replaced by a new generation of Kaolite-based ('Y-12 patented insulation') packages capable of withstanding the dynamic crush test 10 CFR 71.73(c)(2). Three replacement packages are in various stages of development; two are in use. The U.S. Department of Transportation (DOT) 6M specification package, which does not conform to the U.S. Nuclear Regulatory Commission requirements for Type-B packages, is no longer authorized for service on public roads. The ES-3100 shipping package is an example of a Kaolite-based Type-B fissile material package developed as a replacement package for the DOT 6M. With expanded utility, the ES-3100 is designed and licensed for transporting highly enriched uranium and plutonium materials on public roads. The ES-3100 provides added capability for air transport of up to 7-kg quantities of uranium material. This paper presents the productivity techniques and quality aspects in the criticality safety evaluation of Y-12 packages using the ES-3100 as an example.

  8. DOE Cites University of Chicago for Nuclear Safety Violations | Department

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613PortsmouthBartlesvilleAbout »Department of Energy Safety and Ecologyof Energy

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613PortsmouthBartlesvilleAbout »Department of Energy Safety and EcologyofDepartment of

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613PortsmouthBartlesville EnergyDepartment of EnergyCONFIGURATIONNuclear Safety

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

    E-Print Network [OSTI]

    Zelkowitz, Marvin V.

    with requirements for high reliability and mission safety taxes current development methods. In this paper we, Software safety and reliability 1 INTRODUCTION The use of an independent group to provide verification of the process used to ensure mission safety and reliability for NASA Space Shuttle software. The software has

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible for Renewable Energy:Nanowire3627 Federal Register /76Safeguards and SecurityNuclear

  13. Reliability Engineering and System Safety 92 (2007) 15631583 Using fuzzy self-organising maps for safety critical systems

    E-Print Network [OSTI]

    Kelly, Tim

    , the underpinning of the devised model is based upon an existing neuro-fuzzy system called the fuzzy self conventional methods [1] in areas of pattern recognition and function approximation. They are good toolsReliability Engineering and System Safety 92 (2007) 1563­1583 Using fuzzy self-organising maps

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

    SciTech Connect (OSTI)

    none,

    2004-02-28T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Enercon Services, Inc.

    2011-03-14T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    NONE

    1993-09-01T23:59:59.000Z

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

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd of Year 2010Salt |ExelonFAQ: RelocationCriticalityFireof

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

    DOE Patents [OSTI]

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

    1991-01-01T23:59:59.000Z

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

  19. FAQS Reference Guide - Nuclear Safety Specialist | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd of Year 2010SaltInstrumentation and Control FAQS Reference GuideNuclear

  20. Nuclear Explosive Safety Study Functional Area Qualification Standard

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin of Contamination in ManyDepartment of Energy NorthB O|WorkNationalNuclear Energyi NOT

  1. Nuclear Explosives Safety Study Functional Area Qualification Standard

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn'tOrigin of Contamination in ManyDepartment of Energy NorthB O|WorkNationalNuclear

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page onYouTube YouTube Note: Since the.pdfBreaking ofOilNEWResponse to Time-Based Rates fromNuclear Securityand Facility

  3. Safety aspects of ground testing for large nuclear rockets

    SciTech Connect (OSTI)

    Goldman, M.I.

    1988-02-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Broadhead, B.L.

    1991-10-01T23:59:59.000Z

    Currently, UF{sub 6} cylinders designed to contain 2{1/2} tons of UF{sub 6} are classified as Fissile Class 2 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 1 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 {sup 235}U 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 1 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 1 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. 11 refs., 13 figs., 7 tabs.

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 111 1,613PortsmouthBartlesvilleAbout »Department of Energy Safety and Ecology Corp. for

  6. ccsd-00001474,version2-6Sep2004 Model-independent tracking of criticality signals in nuclear multifragmentation data

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    ccsd-00001474,version2-6Sep2004 Model-independent tracking of criticality signals in nuclear-CNRS et Universit´e F-69622 Villeurbanne, France. 8 Institute of Nuclear Physics, Pl-31342 Krak´ow, Poland. 9 National Institute for Physics and Nuclear Engineering, RO-76900 Bucharest-Magurele, Romania

  7. 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.

    1982-03-01T23:59:59.000Z

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

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

    SciTech Connect (OSTI)

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

    2002-02-26T23:59:59.000Z

    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.

  9. 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.

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

    SciTech Connect (OSTI)

    KESSLER, S.F.

    2000-08-10T23:59:59.000Z

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

  11. 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.

    1983-12-01T23:59:59.000Z

    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.

  12. Human Missions to Mars: Designing decision-support tools for a safety critical environment

    E-Print Network [OSTI]

    Whitely, I.; Bogatyreva, O.; Johnson, C.W.

    Whitely,I. Bogatyreva,O. Johnson,C.W. Wolff,M. Townend,M. Proceedings of the 3rd International Association for the Advancement of Space Safety (IAASS) Conference, â??Building a safer space togetherâ??, International Association for the Advancement of Space Safety (IAASS) Rome, Italy

  13. CYBER SECURITY THREATS TO SAFETY-CRITICAL, SPACE-BASED INFRASTRUCTURES

    E-Print Network [OSTI]

    Johnson, Chris

    into advanced air-traffic management applications, rail signalling systems, energy distribution software etc of a combined safety and security risk assessment methodology. The aim is to identify attack scenarios, the inclusion of a combined safety-security risk based methodology for the identification of attack scenarios

  14. Standard practice for qualification and acceptance of boron based metallic neutron absorbers for nuclear criticality control for dry cask storage systems and transportation packaging

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2007-01-01T23:59:59.000Z

    1.1 This practice provides procedures for qualification and acceptance of neutron absorber materials used to provide criticality control by absorbing thermal neutrons in systems designed for nuclear fuel storage, transportation, or both. 1.2 This practice is limited to neutron absorber materials consisting of metal alloys, metal matrix composites (MMCs), and cermets, clad or unclad, containing the neutron absorber boron-10 (10B). 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

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

    SciTech Connect (OSTI)

    Sholtis, J.A. Jr. (Directorate of Nuclear Surety, Air Force Inspection Safety Center, Kirtland AFB, New Mexico 87117-5000 (USA)); Gray, L.B. (Office of Safety Mission Quality, National Aeronautics Space Administration, Washington, DC 20546 (USA)); Huff, D.A. (Office of Environment, Safety, Health, U.S. Department of Energy, Washington, DC 20545 (USA)); Klug, N.P. (Office of Light Water Reactor Safety Technology, U.S. Department of Energy, Washington, DC 20545 (USA)); Winchester, R.O. (Directorate of Nuclear Surety, Air Force Inspection Safety Center, Kirtland, AFB, New Mexico 87111-5000 (USA))

    1991-01-01T23:59:59.000Z

    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.

  16. Observation of the critical end point in the phase diagram for hot and dense nuclear matter

    E-Print Network [OSTI]

    Lacey, Roy A

    2014-01-01T23:59:59.000Z

    Excitation functions for the Gaussian emission source radii difference ($R^2_{\\text{out}} - R^2_{\\text{side}}$) obtained from two-pion interferometry measurements in Au+Au ($\\sqrt{s_{NN}}= 7.7 - 200$ GeV) and Pb+Pb ($\\sqrt{s_{NN}}= 2.76$ TeV) collisions, are studied for a broad range of collision centralities. The observed non-monotonic excitation functions validate the finite-size scaling patterns expected for the deconfinement phase transition and the critical end point (CEP), in the temperature vs. baryon chemical potential ($T,\\mu_B$) plane of the nuclear matter phase diagram. A Finite-Size Scaling (FSS) analysis of these data indicate a second order phase transition with the estimates $T^{\\text{cep}} \\sim 165$~MeV and $\\mu_B^{\\text{cep}} \\sim 100$~MeV for the location of the critical end point. The critical exponents ($\

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

    E-Print Network [OSTI]

    Achorn, Ted Glen

    1991-01-01T23:59:59.000Z

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

  18. 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)]

    1997-12-12T23:59:59.000Z

    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.

  19. Securing the Interface: Safety-Critical Interaction between Humans and Mobile Robots

    E-Print Network [OSTI]

    Ladkin, Peter B.

    -and-safety laws requiring interlocks to prevent people being within the reach- radius of fixed-base industrial robots. And there are people working on principles of engagement for artifical warriors, which we may

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

    E-Print Network [OSTI]

    damage, and release of 1.5M barrels of crude oil and gasoline. The 2002 Pipeline Safety Act was a direct was coated with a corrosion inhibitor. The residual stress led to uneven crack growth through the pipe

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

    E-Print Network [OSTI]

    2004-01-01T23:59:59.000Z

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

  2. Sandia Energy - Risk and Safety Assessment

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

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

  3. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11T23:59:59.000Z

    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.

  4. SciTech Connect: Criticality Safety Basics for INL FMHs and CSOs

    Office of Scientific and Technical Information (OSTI)

    public from the National Technical Information Service, Springfield, VA at www.ntis.gov. Nuclear power is a valuable and efficient energy alternative in our energy-intensive...

  5. SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS Piero Baraldi Chevalier EDF R&D ­ Simulation and information Technologies for Power generation system Department 6, Quai Monitoring, Empirical Modeling, Power Plants, Safety Critical Nuclear Instrumentation, Autoassociative models

  6. Pacific Northwest Laboratory: Annual report for 1986 to the Assistant Secretary for Environment, Safety and Health: Part 5, Nuclear and operational safety

    SciTech Connect (OSTI)

    Faust, L.G.; Kennedy, W.E.; Steelman, B.L.; Selby, J.M.

    1987-02-01T23:59:59.000Z

    Part 5 of the 1986 Annual Report to the Department of Energy's Assistant Secretary for Environment, Safety and Health presents Pacific Northwest Laboratory's progress on work performed for the Office of Nuclear Safety, the Office of Operational Safety, and for the Office of Environmental Analysis. For each project, as identified by the Field Task Proposal/Agreement, articles describe progress made during fiscal year 1986. Authors of these articles represent a broad spectrum of capabilities derived from three of the seven research departments of the Laboratory, reflecting the interdisciplinary nature of the work.

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

    E-Print Network [OSTI]

    The 6th International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6 Junbeom Yoo, Suhyun Park, Hojung Bang, Taihyo Kim, Sungdeok Cha Korea Advanced Institute of Science for assuring quality of software. In the area of nuclear power plant control systems, testing on software

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

    SciTech Connect (OSTI)

    None

    2001-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2014-06-01T23:59:59.000Z

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

  10. Techniques to evaluate the importance of common cause degradation on reliability and safety of nuclear weapons.

    SciTech Connect (OSTI)

    Darby, John L.

    2011-05-01T23:59:59.000Z

    As the nuclear weapon stockpile ages, there is increased concern about common degradation ultimately leading to common cause failure of multiple weapons that could significantly impact reliability or safety. Current acceptable limits for the reliability and safety of a weapon are based on upper limits on the probability of failure of an individual item, assuming that failures among items are independent. We expanded the current acceptable limits to apply to situations with common cause failure. Then, we developed a simple screening process to quickly assess the importance of observed common degradation for both reliability and safety to determine if further action is necessary. The screening process conservatively assumes that common degradation is common cause failure. For a population with between 100 and 5000 items we applied the screening process and conclude the following. In general, for a reliability requirement specified in the Military Characteristics (MCs) for a specific weapon system, common degradation is of concern if more than 100(1-x)% of the weapons are susceptible to common degradation, where x is the required reliability expressed as a fraction. Common degradation is of concern for the safety of a weapon subsystem if more than 0.1% of the population is susceptible to common degradation. Common degradation is of concern for the safety of a weapon component or overall weapon system if two or more components/weapons in the population are susceptible to degradation. Finally, we developed a technique for detailed evaluation of common degradation leading to common cause failure for situations that are determined to be of concern using the screening process. The detailed evaluation requires that best estimates of common cause and independent failure probabilities be produced. Using these techniques, observed common degradation can be evaluated for effects on reliability and safety.

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

    SciTech Connect (OSTI)

    Naus, D.J.; Oland, C.B. (Oak Ridge National Lab., TN (USA)); Arndt, E.G. (Nuclear Regulatory Commission, Washington, DC (USA))

    1990-01-01T23:59:59.000Z

    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.

  12. Criticality Evaluation - Cask Unloading Using the Shielded Transfer System

    SciTech Connect (OSTI)

    Blanchard, A. [Westinghouse Savannah River Company, AIKEN, SC (United States); Nadeau, M.L.

    1998-06-01T23:59:59.000Z

    This evaluation reviewed the criticality aspects of Shielded Transfer System (STS) operation. The existing Nuclear Safety Control (NSC) elements were examined and new NSC elements were developed, as needed.

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

    SciTech Connect (OSTI)

    Gilli, Ludivine; Charron, Sylvie [Institut de Radioprotection et de Surete Nucleaire (IRSN), Fontenay-aux-Roses (France)

    2012-07-01T23:59:59.000Z

    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)

  14. Nuclear critical charge for two-electron ion in Lagrange mesh method

    E-Print Network [OSTI]

    H. Olivares Pilón; A. V. Turbiner

    2014-12-16T23:59:59.000Z

    The Schroedinger equation for two electrons in the field of a charged fixed center $Z$ is solved with the Lagrange mesh method for charges close to the critical charge $Z_{cr}$. We confirm the value of the nuclear critical charge $Z_{cr}$ recently calculated in Estienne et al. {\\em Phys. Rev. Lett. \\bf 112}, 173001 (2014) to 11 decimal digits using an inhomogeneous (non-uniform) three-dimensional lattice of size $70 \\times 70 \\times 20$. We show that the ground state energy for H$^-$ is accurate to 14 decimals on the lattice $50 \\times 50 \\times 40$ in comparison with the highly accurate result by Nakashima-Nakatsuji, {\\it J. Chem. Phys. \\bf 127}, 224104 (2007).

  15. Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel

    SciTech Connect (OSTI)

    L. Angers

    2001-01-31T23:59:59.000Z

    The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

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

    SciTech Connect (OSTI)

    Not Available

    1992-08-01T23:59:59.000Z

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

  17. Institutional implications of establishing safety goals for nuclear power plants. [PWR; BWR

    SciTech Connect (OSTI)

    Morris, F.A.; Hooper, R.L.

    1983-07-01T23:59:59.000Z

    The purpose of this project is to anticipate and address institutional problems that may arise from the adoption of NRC's proposed Policy Statement on Safety Goals for Nuclear Power Plants. The report emphasizes one particular category of institutional problems: the possible use of safety goals as a basis for legal challenges to NRC actions, and the resolution of such challenges by the courts. Three types of legal issues are identified and analyzed. These are, first, general legal issues such as access to the legal system, burden of proof, and standard of proof. Second is the particular formulation of goals. Involved here are such questions as sustainable rationale, definitions, avoided issues, vagueness of time and space details, and degree of conservatism. Implementation brings up the third set of issues which include interpretation and application, linkage to probabilistic risk assessment, consequences as compared to events, and the use of results.

  18. Microbiological assessment of critical control points and safety of precooked beef products

    E-Print Network [OSTI]

    Michel, Michael Ellis

    1990-01-01T23:59:59.000Z

    , Staphylococcus auretts', Escherichia coli, Listeria monocytogenes, and Salmonella typhimurium) inoculated onto precooked, vacuum-packaged, sliced roast beef was determined. Samples consisted of two slices of roast beef that were inoculated and vacuum.... ACKNOWLEDGEMENTS. Vl LIST OF TABLES. . . LIST OF FIGURES. . . INTRODUCTION. REVIEW OF LITERATURE Precooked, vacuum-packaged, cured meat. Precooked, vacuum-packaged, uncured meat Safety of precooked beef . MATERIALS AND METHODS...

  19. Said Mughaghab Nuclear Data 2010,

    E-Print Network [OSTI]

    Danon, Yaron

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

  20. Effect of {gamma}-irradiation on strength of concrete for nuclear-safety structures

    SciTech Connect (OSTI)

    Vodak, F. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic); Trtik, K. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic); Sopko, V. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic); Kapickova, O. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic); Demo, P. [Czech Technical University (CVUT), Faculty of Civil Engineering, Prague, Thakurova 7, CZ 166 29 Prague 6 (Czech Republic)]. E-mail: demo@fzu.cz

    2005-07-01T23:59:59.000Z

    Concrete applied for construction of nuclear power plant (NPP) Temelin (Czech Republic) has been exposed to {gamma}-irradiation up to dose 6x10{sup 5} Gy. Depending on the level of irradiation, changes in strength, porous structure and phase composition of the concrete have been studied. It is found that irradiation lowers both the strength of concrete (about 10%) and volume (resp. surface) of porous space. On the other hand, {gamma}-irradiation increases the ratio of calcite, CaCO{sub 3}, in the concrete. Observed effects are discussed with respect to safety of NPPs.

  1. Critical Drivers for Safety Culture: Examining Department of Energy and U.S. Army Operational Experiences - 12382

    SciTech Connect (OSTI)

    Lowes, Elizabeth A. [The S.M. Stoller Corporation, Broomfield, Colorado (United States)

    2012-07-01T23:59:59.000Z

    Evaluating operational incidents can provide a window into the drivers most critical to establishing and maintaining a strong safety culture, thereby minimizing the potential project risk associated with safety incidents. By examining U.S. Department of Energy (DOE) versus U.S. Army drivers in terms of regulatory and contract requirements, programs implemented to address the requirements, and example case studies of operational events, a view of the elements most critical to making a positive influence on safety culture is presented. Four case studies are used in this evaluation; two from DOE and two from U.S. Army experiences. Although the standards guiding operations at these facilities are different, there are many similarities in the level of hazards, as well as the causes and the potential consequences of the events presented. Two of the incidents examined, one from a DOE operation and the other from a U.S. Army facility, resulted in workers receiving chemical burns. The remaining two incidents are similar in that significant conduct of operations failures occurred resulting in high-level radioactive waste (in the case of the DOE facility) or chemical agent (in the case of the Army facility) being transferred outside of engineering controls. A review of the investigation reports for all four events indicates the primary causes to be failures in work planning leading to ineffective hazard evaluation and control, lack of procedure adherence, and most importantly, lack of management oversight to effectively reinforce expectations for safe work planning and execution. DOE and Army safety programs are similar, and although there are some differences in contractual requirements, the expectations for safe performance are essentially the same. This analysis concludes that instilling a positive safety culture comes down to management leadership and engagement to (1) cultivate an environment that values a questioning attitude and (2) continually reinforce expectations for the appropriate level of rigor in work planning and procedure adherence. A review of the root causes and key contributing causes to the events indicate: - Three of the four root cause analyses cite lack of management engagement (oversight, involvement, ability to recognize issues, etc.) as a root cause to the events. - Two of the four root cause analyses cite work planning failures as a root cause to the events and all cause analyses reflect work planning failures as contributing factors to the events. - All events with the exception of the Tuba City plant shutdown indicate procedure noncompliance as a key contributor; in the case of Tuba City the procedure issues were primarily related to a lack of procedures, or a lack of sufficiently detailed procedures. - All events included discussion or suggestion of a lack of a questioning attitude, either on the part of management/supervision, work planners, or workers. This analysis suggests that the most critical drivers to safety culture are: - Management engagement, - Effective work planning and procedures, and - Procedure adherence with a questioning attitude to ensure procedural problems are identified and fixed. In high-hazard operational environments the importance of robust work planning processes and procedure adherence cannot be overstated. However, having the processes by themselves is not enough. Management must actively engage in expectation setting and ensure work planning that meets expectations for hazard analysis and control, develop a culture that encourages incident reporting and a questioning attitude, and routinely observe work performance to reinforce expectations for adherence to procedures/work control documents. In conclusion, the most critical driver to achieving a workforce culture that supports safe and effective project performance can be summarized as follows: 'Management engagement to continually reinforce expectations for work planning processes and procedure adherence in an environment that cultivates a questioning attitude'. (authors)

  2. Monitoring the Long-Term Safety Performance of a Repository for Used Nuclear Fuel - 12294

    SciTech Connect (OSTI)

    Villagran, J.E. [Nuclear Waste Management Organization, Toronto (Canada)

    2012-07-01T23:59:59.000Z

    The nuclear waste management programs of several nations include plans for the design, construction and operation of deep geological repositories. Some of these programs have initiated the licensing process for their repository designs. Monitoring strategies and systems are at different levels of development in each program and there is common ground with respect to the ultimate goal of the monitoring function. In this context, the primary functions of a monitoring system are considered to be the verification of safety performance and making available information that may be required for implementation of future decisions such as the timing of repository decommissioning and closure or the possible retrieval of waste containers. This study examines some of the relevant issues and outlines a conceptual monitoring system for further study and development during implementation of Adaptive Phased Management, the method selected by the Government of Canada for long-term management of used nuclear fuel. (author)

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

    SciTech Connect (OSTI)

    Not Available

    1990-08-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Uhrig, R.E.

    1988-01-01T23:59:59.000Z

    In the operation of a nuclear power plant, the sheer magnitude of the number of process parameters and systems interactions poses difficulties for the operators, particularly during abnormal or emergency situations. Recovery from an upset situation depends upon the facility with which the available raw data can be converted into and assimilated as meaningful knowledge. Plant personnel are sometimes affected by stress and emotion, which may have varying degrees of influence on their performance. Expert systems can take some of the uncertainty and guesswork out of their decisions by providing expert advice and rapid access to a large information base. Application of artificial intelligence technologies, particularly expert systems, to control room activities in a nuclear power plant has the potential to reduce operator error and improve power plant safety and reliability. 12 refs.

  5. 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.

    1989-01-01T23:59:59.000Z

    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.

  6. Safety of interim storage solutions of used nuclear fuel during extended term

    SciTech Connect (OSTI)

    Shelton, C.; Bader, S.; Issard, H.; Arslan, M. [AREVA, 7135 Minstrel Way, Suite 300 Columbia, MD 21045 (United States)

    2013-07-01T23:59:59.000Z

    In 2013, the total amount of stored used nuclear fuel (UNF) in the world will reach 225,000 T HM. The UNF inventory in wet storage will take up over 80% of the available total spent fuel pool (SFP) capacity. Interim storage solutions are needed. They give flexibility to the nuclear operators and ensure that nuclear reactors continue to operate. However, we need to keep in mind that they are also an easy way to differ final decision and implementation of a UNF management approach (recycling or final disposal). In term of public perception, they can have a negative impact overtime as it may appear that nuclear industry may have significant issues to resolve. In countries lacking an integrated UNF management approach, the UNF are being discharged from the SFPs to interim storage (mostly to dry storage) at the same rate as UNF is being discharged from reactors, as the SFPs at the reactor sites are becoming full. This is now the case in USA, Taiwan, Switzerland, Spain, South Africa and Germany. For interim storage, AREVA has developed different solutions in order to allow the continued operation of reactors while meeting the current requirements of Safety Authorities: -) Dry storage canisters on pads, -) Dual-purpose casks (dry storage and transportation), -) Vault dry storage, and -) Centralized pool storage.

  7. Search for the QCD critical point in nuclear collisions at the CERN SPS

    E-Print Network [OSTI]

    The NA49 Collaboration; N. G. Antoniou; F. K. Diakonos; G. Mavromanolakis

    2010-05-19T23:59:59.000Z

    Pion production in nuclear collisions at the SPS is investigated with the aim to search, in a restricted domain of the phase diagram, for power-laws in the behavior of correlations which are compatible with critical QCD. We have analyzed interactions of nuclei of different size (p+p, C+C, Si+Si, Pb+Pb) at 158$A$ GeV adopting, as appropriate observables, scaled factorial moments in a search for intermittent fluctuations in transverse dimensions. The analysis is performed for $\\pi^+\\pi^-$ pairs with invariant mass very close to the two-pion threshold. In this sector one may capture critical fluctuations of the sigma component in a hadronic medium, even if the $\\sigma$-meson has no well defined vacuum state. It turns out that for the Pb+Pb system the proposed analysis technique cannot be applied without entering the invariant mass region with strong Coulomb correlations. As a result the treatment becomes inconclusive in this case. Our results for the other systems indicate the presence of power-law fluctuations in the freeze-out state of Si+Si approaching in size the prediction of critical QCD.

  8. 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...

  9. Evaluation of natural phenomena hazards as part of safety assessments for nuclear facilities

    SciTech Connect (OSTI)

    Kot, C.A.; Hsieh, B.J.; Srinivasan, M.G.; Shin, Y.W.

    1995-02-01T23:59:59.000Z

    The continued operation of existing US Department of Energy (DOE) nuclear facilities and laboratories requires a safety reassessment based on current criteria and guidelines. This also includes evaluations for the effects of Natural Phenomena Hazards (NPH), for which these facilities may not have been designed. The NPH evaluations follow the requirements of DOE Order 5480.28, Natural Phenomena Hazards Mitigation (1993) which establishes NPH Performance Categories (PCs) for DOE facilities and associated target probabilistic performance goals. These goals are expressed as the mean annual probability of exceedance of acceptable behavior for structures, systems and components (SSCs) subjected to NPH effects. The assignment of an NPH Performance Category is based on the overall hazard categorization (low, moderate, high) of a facility and on the function of an SSC under evaluation (DOE-STD-1021, 1992). Detailed guidance for the NPH analysis and evaluation criteria are also provided (DOE-STD-1020, 1994). These analyses can be very resource intensive, and may not be necessary for the evaluation of all SSCs in existing facilities, in particular for low hazard category facilities. An approach relying heavily on screening inspections, engineering judgment and use of NPH experience data (S. J. Eder et al., 1993), can minimize the analytical effort, give reasonable estimates of the NPH susceptibilities, and yield adequate information for an overall safety evaluation of the facility. In the following sections this approach is described in more detail and is illustrated by an application to a nuclear laboratory complex.

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

    SciTech Connect (OSTI)

    O, J.M.; Higgins, J.; Stephen Fleger - NRC

    2011-09-19T23:59:59.000Z

    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.

  11. Critical Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel Based on Diffusion Controlled Cavity Growth

    SciTech Connect (OSTI)

    Hayes, T.A.; Rosen, R.S.; Kassner, M.E.

    1999-12-01T23:59:59.000Z

    Interim dry storage of spent nuclear fuel (SNF) rods is of critical concern because a shortage of existing SNF wet storage capacity combined with delays in the availability of a permanent disposal repository has led to an increasing number of SNF rods being placed into interim dry storage. Safe interim dry storage must be maintained for a minimum of twenty years according to the Standard Review Plan for Dry Cask Storage Systems [1] and the Code of Federal Regulations, 10 CFR Part 72 [2]. Interim dry storage licensees must meet certain safety conditions when storing SNF rods to ensure that there is a ''very low probability (e.g. 0.5%) of cladding breach during long-term storage'' [1]. Commercial SNF typically consists of uranium oxide pellets surrounded by a thin cladding. The cladding is usually an {alpha}-zirconium based alloy know as ''Zircaloy''. In dry storage, the SNF rods are confined in one of several types of cask systems approved by the Nuclear Regulatory Commission (NRC). ''The cask system must be designed to prevent degradation of fuel cladding that results in a type of cladding breach, such as axial-splits or ductile fracture, where irradiated UO{sub 2} particles may be released. In addition, the fuel cladding should not degrade to the point where more than one percent of the fuel rods suffer pinhole or hairline crack type failure under normal storage conditions [1].'' The NRC has approved two models [3,4] for use by proposed dry storage licensees to determine the maximum initial temperature limit for nuclear fuel rods in dry storage that supposedly meet the above criteria and yield consistent temperature limits. Though these two models are based on the same fundamental failure theory, different assumptions have been made including the choice of values for material constants in the failure equation. This report will examine and compare the similarities and inconsistencies of these two models. It will illustrate some of the shortcomings of the current models and suggest modifications as well as some experiments that should be started in the near future. This report will also discuss changes in the current NRC standards with regard to the adoption of a strain-based model to be used to determine maximum allowable temperatures of the SNF.

  12. Criticality Safety Scoping Study for the Transport of Weapons-Grade Mixed-Oxide Fuel Using the MO-1 Shipping Package

    SciTech Connect (OSTI)

    Dunn, M.E.; Fox, P.B.

    1999-05-01T23:59:59.000Z

    This report provides the criticality safety information needed for obtaining certification of the shipment of mixed-oxide (MOX) fuel using the MO-1 [USA/9069/B()F] shipping package. Specifically, this report addresses the shipment of non-weapons-grade MOX fuel as certified under Certificate of Compliance 9069, Revision 10. The report further addresses the shipment of weapons-grade MOX fuel using a possible Westinghouse fuel design. Criticality safety analysis information is provided to demonstrate that the requirements of 10 CFR S 71.55 and 71.59 are satisfied for the MO-1 package. Using NUREG/CR-5661 as a guide, a transport index (TI) for criticality control is determined for the shipment of non-weapons-grade MOX fuel as specified in Certificate of Compliance 9069, Revision 10. A TI for criticality control is also determined for the shipment of weapons-grade MOX fuel. Since the possible weapons-grade fuel design is preliminary in nature, this report is considered to be a scoping evaluation and is not intended as a substitute for the final criticality safety analysis of the MO-1 shipping package. However, the criticality safety evaluation information that is presented in this report does demonstrate the feasibility of obtaining certification for the transport of weapons-grade MOX lead test fuel using the MO-1 shipping package.

  13. A Critical Review of Practice of Equating the Reactivity of Spent Fuel to Fresh Fuel in Burnup Credit Criticality Safety Analyses for PWR Spent Fuel Pool Storage

    SciTech Connect (OSTI)

    Wagner, J.C.; Parks, C.V.

    2000-09-01T23:59:59.000Z

    This research examines the practice of equating the reactivity of spent fuel to that of fresh fuel for the purpose of performing burnup credit criticality safety analyses for PWR spent fuel pool (SFP) storage conditions. The investigation consists of comparing k{sub inf} estimates based on reactivity equivalent fresh fuel enrichment (REFFE) to k{sub inf} estimates using the actual spent fuel isotopics. Analyses of selected storage configurations common in PWR SFPs show that this practice yields nonconservative results (on the order of a few tenths of a percent) in configurations in which the spent fuel is adjacent to higher-reactivity assemblies (e.g., fresh or lower-burned assemblies) and yields conservative results in configurations in which spent fuel is adjacent to lower-reactivity assemblies (e.g., higher-burned fuel or empty cells). When the REFFE is determined based on unborated water moderation, analyses for storage conditions with soluble boron present reveal significant nonconservative results associated with the use of the REFFE. This observation is considered to be important, especially considering the recent allowance of credit for soluble boron up to 5% in reactivity. Finally, it is shown that the practice of equating the reactivity of spent fuel to fresh fuel is acceptable, provided the conditions for which the REFFE was determined remain unchanged. Determination of the REFFE for a reference configuration and subsequent use of the REFFE for different configurations violates the basis used for the determination of the REFFE and, thus, may lead to inaccurate, and possibly, nonconservative estimates of reactivity. A significant concentration ({approximately}2000 ppm) of soluble boron is typically (but not necessarily required to be) present in PWR SFPs, of which only a portion ({le} 500 ppm) may be credited in safety analyses. Thus, a large subcritical margin currently exists that more than accounts for errors or uncertainties associated with the use of the REFFE. Consequently, the findings presented here do not represent a significant safety concern unless/until the subcritical margin associated with the soluble boron (that is not currently explicitly credited) is offset by the uncertainties associated with burnup credit and/or the expanded allowance of credit for the soluble boron.

  14. Application of Diagnostic/Prognostic Methods to Critical Equipment for the Spent Nuclear Fuel Cleanup Program

    SciTech Connect (OSTI)

    Casazza, Lawrence O.; Jarrell, Donald B.; Koehler, Theresa M.; Meador, Richard J.; Wallace, Dale E.

    2002-02-28T23:59:59.000Z

    The management of the Spent Nuclear Fuel (SNF) project at the Hanford K-Basin in the 100 N Area has successfully restructured the preventive maintenance, spare parts inventory requirements, and the operator rounds data requirements. In this investigation, they continue to examine the different facets of the operations and maintenance (O&M) of the K-Basin cleanup project in search of additional reliability and cost savings. This report focuses on the initial findings of a team of PNNL engineers engaged to identify potential opportunities for reducing the cost of O&M through the application of advanced diagnostics (fault determination) and prognostics (residual life/reliability determination). The objective is to introduce predictive technologies to eliminate or reduce high impact equipment failures. The PNNL team in conjunction with the SNF engineers found the following major opportunities for cost reduction and/or enhancing reliability: (1) Provide data routing and automated analysis from existing detection systems to a display center that will engage the operations and engineering team. This display will be operator intuitive with system alarms and integrated diagnostic capability. (2) Change operating methods to reduce major transients induced in critical equipment. This would reduce stress levels on critical equipment. (3) Install a limited sensor set on failure prone critical equipment to allow degradation or stressor levels to be monitored and alarmed. This would provide operators and engineers with advance guidance and warning of failure events. Specific methods for implementation of the above improvement opportunities are provided in the recommendations. They include an Integrated Water Treatment System (IWTS) decision support system, introduction of variable frequency drives on certain pump motors, and the addition of limited diagnostic instrumentation on specified critical equipment.

  15. 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. [Los Alamos National Laboratory

    2012-06-19T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    None

    2000-02-01T23:59:59.000Z

    This is the tenth Annual Report to the Congress describing Department of Energy activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board (Board). The Board, an independent executive-branch agency established in 1988, provides advice and recommendations to the Secretary of Energy regarding public health and safety issues at the Department's defense nuclear facilities. The Board also reviews and evaluates the content and implementation of health and safety standards, as well as other requirements, relating to the design, construction, operation, and decommissioning of the Department's defense nuclear facilities. During 1999, Departmental activities resulted in the closure of nine Board recommendations. In addition, the Department has completed all implementation plan milestones associated with three Board recommendations. One new Board recommendation was received and accepted by the Department in 1999, and a new implementation plan is being developed to address this recommendation. The Department has also made significant progress with a number of broad-based initiatives to improve safety. These include expanded implementation of integrated safety management at field sites, opening of a repository for long-term storage of transuranic wastes, and continued progress on stabilizing excess nuclear materials to achieve significant risk reduction.

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

    SciTech Connect (OSTI)

    NONE

    1999-02-01T23:59:59.000Z

    This is the ninth Annual Report to the Congress describing Department of Energy (Department) activities in response to formal recommendations and other interactions with the Defense Nuclear Facilities Safety Board (Board). The Board, an independent executive-branch agency established in 1988, provides advice and recommendations to the Secretary of energy regarding public health and safety issues at the Department`s defense nuclear facilities. The Board also reviews and evaluates the content and implementation of health and safety standards, as well as other requirements, relating to the design, construction, operation, and decommissioning of the Department`s defense nuclear facilities. The locations of the major Department facilities are provided. During 1998, Departmental activities resulted in the proposed closure of one Board recommendation. In addition, the Department has completed all implementation plan milestones associated with four other Board recommendations. Two new Board recommendations were received and accepted by the Department in 1998, and two new implementation plans are being developed to address these recommendations. The Department has also made significant progress with a number of broad-based initiatives to improve safety. These include expanded implementation of integrated safety management at field sites, a renewed effort to increase the technical capabilities of the federal workforce, and a revised plan for stabilizing excess nuclear materials to achieve significant risk reduction.

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

    SciTech Connect (OSTI)

    Wittenbrock, N. G.

    1982-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Yi Zhang; Michael W. Golay

    2003-10-01T23:59:59.000Z

    The work of our project is intended to help introducing digital technologies into nuclear power into nuclear power plant safety related software applications. In our project we utilize a combination of modern software engineering methods: design process discipline and feedback, formal methods, automated computer aided software engineering tools, automatic code generation, and extensive feasible structure flow path testing to improve software quality. The tactics include ensuring that the software structure is kept simple, permitting routine testing during design development, permitting extensive finished product testing in the input data space of most likely service and using test-based Bayesian updating to estimate the probability that a random software input will encounter an error upon execution. From the results obtained the software reliability can be both improved and its value estimated. Hopefully our success in the project's work can aid the transition of the nuclear enterprise into the modern information world. In our work, we have been using the proprietary sample software, the digital Signal Validation Algorithm (SVA), provided by Westinghouse. Also our work is being done with their collaboration. The SVA software is used for selecting the plant instrumentation signal set which is to be used as the input the digital Plant Protection System (PPS). This is the system that automatically decides whether to trip the reactor. In our work, we are using -001 computer assisted software engineering (CASE) tool of Hamilton Technologies Inc. This tool is capable of stating the syntactic structure of a program reflecting its state requirements, logical functions and data structure.

  20. Exploration of high-dimensional scalar function for nuclear reactor safety analysis and visualization

    SciTech Connect (OSTI)

    Maljovec, D.; Wang, B.; Pascucci, V. [Scientific Computing and Imaging Institute, University of Utah (United States); Bremer, P. T. [Lawrence Livermore National Laboratory (United States); Pernice, M.; Mandelli, D.; Nourgaliev, R. [Idaho National Laboratory (United States)

    2013-07-01T23:59:59.000Z

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user's guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving data from a nuclear reactor safety simulation. (authors)

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

    SciTech Connect (OSTI)

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

    2013-05-01T23:59:59.000Z

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

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

    SciTech Connect (OSTI)

    Uhrig, R.E.

    1987-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1993-06-01T23:59:59.000Z

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

  4. Criticality Safety Evaluation for TRU Waste In Storage at the RWMC

    SciTech Connect (OSTI)

    M. E. Shaw; J. B. Briggs; C. A. Atkinson; G. J. Briscoe

    1994-04-01T23:59:59.000Z

    Stored containers (drums, boxes, and bins) of transuranic waste at the Radioactive Waste Management Complex (RWMC) facility located at the Idaho National Engineering Laboratory (INEL) were evaluated based on inherent neutron absorption characteristics of the waste materials. It was demonstrated that these properties are sufficient to preclude a criticality accident at the actual fissile levels present in the waste stored at the RWMC. Based on the database information available, the results reported herein confirm that the waste drums, boxes, and bins currently stored at the RWMC will remain safely subcritical if rearranged, restacked, or otherwise handled. Acceptance criteria for receiving future drum shipments were established based on fully infinite systems.

  5. Benchmark calculations on the phase II problem of uncertainty analyses for criticality safety assessment

    SciTech Connect (OSTI)

    Lee, G. S.; Lee, J.; Kim, G. Y.; Woo, S. W. [Korea Inst. of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon, 305-338 (Korea, Republic of)

    2012-07-01T23:59:59.000Z

    The phase II benchmark problem of expert group UACSA includes a configuration of a PWR fuel storage rack and focuses on the uncertainty of criticality from manufacturing tolerance of design parameters such as fuel enrichment, density, diameter, thickness of neutron absorber and structural material, and so on. It provides probability density functions for each design parameter. In this paper, upper limits of k{sub eff} of 95%/95% tolerance with two methods are calculated by sampling design parameters using given probability distributions and compared with the result from traditional approach. (authors)

  6. Safety research programs sponsored by Office of Nuclear Regulatory Research: Progress report, January 1--March 31, 1989

    SciTech Connect (OSTI)

    Weiss, A.J. (comp.)

    1989-08-01T23:59:59.000Z

    This progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Regulatory Applications, Division of Engineering, Division of Safety Issue Resolution, and Division of Systems Research of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research following the reorganization in July 1988. The previous reports have covered the period October 1, 1976 through December 31, 1988.

  7. International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9) N9P0302 Kaohsiung, Taiwan, September 9-13, 2012

    E-Print Network [OSTI]

    Haviland, David

    The 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9 Kudinov Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, SE proportional to terminal melt spread thickness. At certain thickness, the melt layer becomes non

  8. CSRL-V: an ENDF/B-V 227-group cross section library for criticality safety studies

    SciTech Connect (OSTI)

    Ford, W.E. III; Westfall, R.M.; Diggs, B.R.; Webster, C.C.; Knight, J.R.

    1980-01-01T23:59:59.000Z

    The AMPX system was used to generate a P/sub 3/ 227-neutron-group master cross-section library containing data for all materials in the ENDF/B-V general purpose file. CSRL-V is a data base for the subsequent generation of problem-dependent fine- and/or broad-group cross sections for shipping cask calculations and other criticality safety analyses. The problem-dependent data can be used with codes such as KENO IV, ANISN, XSDRNPM, VENTURE, DOT, MORSE, etc. CSRL-V data can be coupled with photon-production and photon-interaction multigroup data produced with the AMPX system to produce coupled neutron-gamma cross-section libraries. Consideration was given to the resonance structure of prominent nuclei, the thresholds of important reactions, and various fission spectra. Data in the CSRL-V library were checked for first-order consistencies and tested in performance parameter calculations for a series of benchmark critical experiments. The CSRL-V library is available on magnetic tape. 1 table. (RWR)

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

    SciTech Connect (OSTI)

    Barnett, C.S.

    1985-11-01T23:59:59.000Z

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

  10. The safety review and approval process for space nuclear power sources

    SciTech Connect (OSTI)

    Bennett, G.L. [National Aeronautics and Space Administration, Washington, DC (United States)

    1991-01-01T23:59:59.000Z

    Over the past 30 yr. the U.S. Government has evolved a process for the safety review and launch approval of nuclear power sources (NPSs) proposed for launch into space. This process, which involves a number of governmental agencies, ensures that the various postulated accident scenarios are considered, that the responses of the NPSs to the accident environments are assessed, and that appropriate elements of the Federal Government are involved in the launch approval. This process has worked very well in the successful launches of 37 radioisotope thermoelectric generators and 1 reactor by the United States since 1961. Particular attention will be focused on the recent launch of the Galileo spacecraft. 19 refs., 12 figs., 4 tabs.

  11. Safety research programs sponsored by Office of Nuclear Regulatory Research: Quarterly progress report, July 1-September 30, 1986

    SciTech Connect (OSTI)

    Bari, R.A.; Bezler, P.; Boccio, J.L.; Ginsberg, T.; Greene, G.A.; Guppy, J.G.; Hall, R.E.; Hofmayer, C.H.; Khatib-Rahbar, H.; Luckas, W.J. Jr.

    1987-03-01T23:59:59.000Z

    This progress report will describe current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Accident Evaluation, Division of Engineering Technology, and Division of Risk Analysis and Operations of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The projects reported are the following: High Temperature Reactor Research, SSC code improvements, Thermal-Hydraulic Reactor Safety Experiments, Thermodynamic Core-Concrete Interaction Experiments and Analysis, Plant Analyzer, Code Assessment and Application, Code Maintenance (RAMONA-3B), MELCOR Verification and Benchmarking, Source Term Code Package Verification and Benchmarking, Uncertainty Analysis of the Source Term; Stress Corrosion Cracking of PWR Steam Generator Tubing, Soil-Structure Interaction Evaluation and Structural Benchmarks, Identification of Age Related Failure Modes; Application of HRA/PRA Results to Support Resolution of Generic Safety Issues Involving Human Performance, Protective Action Decisionmaking, Rebaseling of Risk for Zion, Containment Performance Design Objective, and Operational Safety Reliability Research.

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

    SciTech Connect (OSTI)

    Biswas, D; Mennerdahl, D

    2008-06-23T23:59:59.000Z

    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.

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742Energy ChinaofSchaefer To: CongestionDevelopment ofofthePerformance RequirementsSTD-3007-2007

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't Your Destiny:Revised Finding of No Significant Impact610-9464-941 February 2005Changes

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

    DOE Patents [OSTI]

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

    1986-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Barnett, C.S.

    1985-08-20T23:59:59.000Z

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

  17. Technical basis for environmental qualification of microprocessor-based safety-related equipment in nuclear power plants

    SciTech Connect (OSTI)

    Korsah, K.; Wood, R.T. [Oak Ridge National Lab., TN (United States); Hassan, M. [Brookhaven National Lab., Upton, NY (United States); Tanaka, T.J. [Sandia National Labs., Albuquerque, NM (United States)

    1998-01-01T23:59:59.000Z

    This document presents the results of studies sponsored by the Nuclear Regulatory Commission (NRC) to provide the technical basis for environmental qualification of computer-based safety equipment in nuclear power plants. The studies were conducted by Oak Ridge National Laboratory (ORNL), Sandia National Laboratories (SNL), and Brookhaven National Laboratory (BNL). The studies address the following: (1) adequacy of the present test methods for qualification of digital I and C systems; (2) preferred (i.e., Regulatory Guide-endorsed) standards; (3) recommended stressors to be included in the qualification process during type testing; (4) resolution of need for accelerated aging for equipment to be located in a benign environment; and (5) determination of an appropriate approach for addressing the impact of smoke in digital equipment qualification programs. Significant findings from the studies form the technical basis for a recommended approach to the environmental qualification of microprocessor-based safety-related equipment in nuclear power plants.

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

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

  20. Expectations on Documented Safety Analysis for Deactivated Inactive Nuclear Facilities in a State of Long Term Surveillance & Maintenance or Decommissioning

    SciTech Connect (OSTI)

    JACKSON, M.W.

    2002-05-01T23:59:59.000Z

    DOE promulgated 10 CFR 830 ''Nuclear Safety Management'' on October 10, 2000. Section 204 of the Rule requires that contractors at DOE hazard category 1, 2, and 3 nuclear facilities develop a ''Documented Safety Analysis'' (DSA) that summarizes the work to be performed, the associated hazards, and hazard controls necessary to protect workers, the public, and the environment. Table 2 of Appendix A to the rule has been provided to ensure that DSAs are prepared in accordance with one of the available predetermined ''safe harbor'' approaches. The table presents various acceptable safe harbor DSAs for different nuclear facility operations ranging from nuclear reactors to decommissioning activities. The safe harbor permitted for decommissioning of a nuclear facility encompasses methods described in DOE-STD-1 120-98, ''Integration of Environment, Safety and Health into Facility Disposition Activities,'' and provisions in 29 CFR 1910.120 or 29 CFR 1926.65 (HAZWOPER). Additionally, an evaluation of public safety impacts and development of necessary controls is required when the facility being decommissioned contains radiological inventory or contamination exceeding the Rule's definition for low-level residual fixed radioactivity. This document discusses a cost-effective DSA approach that is based on the concepts of DOE-STD-I 120 and meets the 10 CFR 830 safe harbor requirements for both transition surveillance and maintenance as well as decommissioning. This DSA approach provides continuity for inactive Hanford nuclear facilities that will eventually transition into decommissioning. It also uses a graded approach that meets the expectations of DOE-STD-3011 and addresses HAZWOPER requirements to provide a sound basis for worker protection, particularly where intrusive work is being conducted.

  1. Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository--BMT1 of the DECOVALEX III project. Part 1: Conceptualization

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste on the safety of a hypothetical nuclear waste repository at the near-field and are presented in three on the safety of nuclear waste repositories. To achieve the second objective, hypothetical benchmark test

  2. Safety

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch > TheNuclear Press ReleasesIn theTreatmentSRS Economic0ScienceSTEMSULI

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

    E-Print Network [OSTI]

    Tetreault-Friend, Melanie

    2014-01-01T23:59:59.000Z

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

  4. 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...

  5. A RE-INTRODUCTION TO ANOMALIES OF CRITICALITY

    SciTech Connect (OSTI)

    PUIGH RJ

    2009-09-09T23:59:59.000Z

    In 1974, a small innocuous document was submitted to the American Nuclear Society's Criticality Safety Division for publication that would have lasting impacts on this nuclear field The author was Duane Clayton, manager of the Battelle Pacific Northwest National Laboratory's Critical Mass Lab, the world's preeminent reactor critical experimenter with plutonium solutions. The document was entitled, 'Anomalies of Criticality'. 'Anomalies...' was a compilation of more than thirty separate and distinct examples of departures from what might be commonly expected in the field of nuclear criticality. Mr. Clayton's publication was the derivative of more than ten thousand experiments and countless analytical studies conducted world-wide on every conceivable reactor system imaginable: from fissile bearing solutions to solids, blocks to arrays of fuel rods, low-enriched uranium oxide systems to pure plutonium and highly enriched uranium systems. After publication, the document was commonly used within the nuclear fuel cycle and reactor community to train potential criticality/reactor analysts, experimenters and fuel handlers on important things for consideration when designing systems with critically 'safe' parameters in mind The purpose of this paper is to re-introduce 'Anomalies of Criticality' to the current Criticality Safety community and to add new 'anomalies' to the existing compendium. By so doing, it is the authors' hope that a new generation of nuclear workers and criticality engineers will benefit from its content and might continue to build upon this work in support of the nuclear renaissance that is about to occur.

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

    SciTech Connect (OSTI)

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

    1993-03-01T23:59:59.000Z

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

  7. 1Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory Nuclear Data Research at RPI

    E-Print Network [OSTI]

    Danon, Yaron

    1Mechanical, Aerospace and Nuclear Engineering nacThe Gaerttner Laboratory Nuclear Data Research at RPI Y. Danon Rensselaer Polytechnic Institute, Troy, NY, 12180 RPI Nuclear Data (RND) 2011 Symposium for Criticality Safety and Reactor Applications Rensselaer Polytechnic Institute, April 27, 2011 #12;2Mechanical

  8. Operational safety enhancement of Soviet-designed nuclear reactors via development of nuclear power plant simulators and transfer of related technology

    SciTech Connect (OSTI)

    Kohut, P.; Epel, L.G.; Tutu, N.K. [and others

    1998-08-01T23:59:59.000Z

    The US Department of Energy (DOE), under the US government`s International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper.

  9. Neutronics for critical fission reactors and subcritical fission in hybrids

    SciTech Connect (OSTI)

    Salvatores, Massimo [CEA-Cadarache, DEN-Dir, Bat. 101, St-Paul-Lez-Durance 13108 (France)

    2012-06-19T23:59:59.000Z

    The requirements of future innovative nuclear fuel cycles will focus on safety, sustainability and radioactive waste minimization. Critical fast neutron reactors and sub-critical, external source driven systems (accelerator driven and fusion-fission hybrids) have a potential role to meet these requirements in view of their physics characteristics. This paper provides a short introduction to these features.

  10. Roadmap to an Engineering-Scale Nuclear Fuel Performance & Safety Code

    SciTech Connect (OSTI)

    Turner, John A [ORNL; Clarno, Kevin T [ORNL; Hansen, Glen A [ORNL

    2009-09-01T23:59:59.000Z

    Developing new fuels and qualifying them for large-scale deployment in power reactors is a lengthy and expensive process, typically spanning a period of two decades from concept to licensing. Nuclear fuel designers serve an indispensable role in the process, at the initial exploratory phase as well as in analysis of the testing results. In recent years fuel performance capabilities based on first principles have been playing more of a role in what has traditionally been an empirically dominated process. Nonetheless, nuclear fuel behavior is based on the interaction of multiple complex phenomena, and recent evolutionary approaches are being applied more on a phenomenon-by-phenomenon basis, targeting localized problems, as opposed to a systematic approach based on a fundamental understanding of all interacting parameters. Advanced nuclear fuels are generally more complex, and less understood, than the traditional fuels used in existing reactors (ceramic UO{sub 2} with burnable poisons and other minor additives). The added challenges are primarily caused by a less complete empirical database and, in the case of recycled fuel, the inherent variability in fuel compositions. It is clear that using the traditional approach to develop and qualify fuels over the entire range of variables pertinent to the U.S. Department of Energy (DOE) Office of Nuclear Energy on a timely basis with available funds would be very challenging, if not impossible. As a result the DOE Office of Nuclear Energy has launched the Nuclear Energy Advanced Modeling and Simulation (NEAMS) approach to revolutionize fuel development. This new approach is predicated upon transferring the recent advances in computational sciences and computer technologies into the fuel development program. The effort will couple computational science with recent advances in the fundamental understanding of physical phenomena through ab initio modeling and targeted phenomenological testing to leapfrog many fuel-development activities. Realizing the full benefits of this approach will likely take some time. However, it is important that the developmental activities for modeling and simulation be tightly coupled with the experimental activities to maximize feedback effects and accelerate both the experimental and analytical elements of the program toward a common objective. The close integration of modeling and simulation and experimental activities is key to developing a useful fuel performance simulation capability, providing a validated design and analysis tool, and understanding the uncertainties within the models and design process. The efforts of this project are integrally connected to the Transmutation Fuels Campaign (TFC), which maintains as a primary objective to formulate, fabricate, and qualify a transuranic-based fuel with added minor actinides for use in future fast reactors. Additional details of the TFC scope can be found in the Transmutation Fuels Campaign Execution Plan. This project is an integral component of the TFC modeling and simulation effort, and this multiyear plan borrowed liberally from the Transmutation Fuels Campaign Modeling and Simulation Roadmap. This document provides the multiyear staged development plan to develop a continuum-level Integrated Performance and Safety Code (IPSC) to predict the behavior of the fuel and cladding during normal reactor operations and anticipated transients up to the point of clad breach.

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

    SciTech Connect (OSTI)

    CARTER, R.P.

    1999-11-19T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Schultz, Peter Andrew

    2011-12-01T23:59:59.000Z

    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.

  13. Computer code for space-time diagnostics of nuclear safety parameters

    SciTech Connect (OSTI)

    Solovyev, D. A.; Semenov, A. A.; Gruzdov, F. V.; Druzhaev, A. A.; Shchukin, N. V.; Dolgenko, S. G.; Solovyeva, I. V.; Ovchinnikova, E. A. [National Research Nuclear Univ. MEPhI, Kashirskoe, 31, 115409, Moscow (Russian Federation)

    2012-07-01T23:59:59.000Z

    The computer code ECRAN 3D (Experimental and Calculation Reactor Analysis) is designed for continuous monitoring and diagnostics of reactor cores and databases for RBMK-1000 on the basis of analytical methods for the interrelation parameters of nuclear safety. The code algorithms are based on the analysis of deviations between the physically obtained figures and the results of neutron-physical and thermal-hydraulic calculations. Discrepancies between the measured and calculated signals are equivalent to obtaining inadequacy between performance of the physical device and its simulator. The diagnostics system can solve the following problems: identification of facts and time for inconsistent results, localization of failures, identification and quantification of the causes for inconsistencies. These problems can be effectively solved only when the computer code is working in a real-time mode. This leads to increasing requirements for a higher code performance. As false operations can lead to significant economic losses, the diagnostics system must be based on the certified software tools. POLARIS, version 4.2.1 is used for the neutron-physical calculation in the computer code ECRAN 3D. (authors)

  14. Aerosol particle transport modeling for preclosure safety studies of nuclear waste repositories

    SciTech Connect (OSTI)

    Gelbard, F. [Sandia National Labs., Albuquerque, NM (USA)

    1989-01-01T23:59:59.000Z

    An important concern for preclosure safety analysis of a nuclear waste repository is the potential release to the environment of respirable aerosol particles. Such particles, less than 10 {mu}m in aerodynamic diameter, may have significant adverse health effects if inhaled. To assess the potential health effects of these particles, it is not sufficient to determine the mass fraction of respirable aerosol. The chemical composition of the particles is also of importance since different radionuclides may pose vastly different health hazards. Thus, models are needed to determine under normal and accident conditions the particle size and the chemical composition distributions of aerosol particles as a function of time and of position in the repository. In this work a multicomponent sectional aerosol model is used to determine the aerosol particle size and composition distributions in the repository. A range of aerosol mass releases with varying mean particle sizes and chemical compositions is used to demonstrate the sensitivities and uncertainties of the model. Decontamination factors for some locations in the repository are presented. 8 refs., 1 tab.

  15. Assessing conformance to safety goals using nonparametric empirical Bayes methods: A nuclear reactor application

    SciTech Connect (OSTI)

    Martz, H.F.; Johnson, J.W. [Los Alamos National Lab., NM (United States)

    1997-01-01T23:59:59.000Z

    Nonparametric empirical Bayes methods are used to develop decision criteria for use in deciding whether the risk of a given facility is compatible with a corresponding specified quantitative safety goal. The criteria utilize the uncertain results of a probabilistic risk assessment (PRA) and are derived from an empirical Bayes point of view in which the results from a set of similar facilities are used to estimate the population variability curve (PVC) for the parameter of interest. The PVC is estimated nonparametrically in the sense that the distributional family to which the PVC belongs is completely unknown and unspecified. For the assumed model, the method guarantees that all facilities ultimately accepted as being compatible with the goal have a prespecified exact assurance probability that the goal is not exceeded. The method also accounts for two possible biases in the PRA results. Criteria are developed for use in assessing the compatibility of nuclear power plant PRA-produced severe core damage frequency estimates with a corresponding subsidiary objective.

  16. 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

    2006-09-01T23:59:59.000Z

    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

  17. Risk-informed incident management for nuclear power plants

    E-Print Network [OSTI]

    Smith, Curtis Lee, 1966-

    2002-01-01T23:59:59.000Z

    Decision making as a part of nuclear power plant operations is a critical, but common, task. Plant management is forced to make decisions that may have safety and economic consequences. Formal decision theory offers the ...

  18. 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

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

    E-Print Network [OSTI]

    Galvez, Cristhian

    2011-01-01T23:59:59.000Z

    Advisory Committee and Generation IV International Forum.Nuclear Energy Agency The Generation IV International Forum.Technology Roadmap for Generation IV Nuclear Energy Systems.

  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. System Design and the Safety Basis

    SciTech Connect (OSTI)

    Ellingson, Darrel

    2008-05-06T23:59:59.000Z

    The objective of this paper is to present the Bechtel Jacobs Company, LLC (BJC) Lessons Learned for system design as it relates to safety basis documentation. BJC has had to reconcile incomplete or outdated system description information with current facility safety basis for a number of situations in recent months. This paper has relevance in multiple topical areas including documented safety analysis, decontamination & decommissioning (D&D), safety basis (SB) implementation, safety and design integration, potential inadequacy of the safety analysis (PISA), technical safety requirements (TSR), and unreviewed safety questions. BJC learned that nuclear safety compliance relies on adequate and well documented system design information. A number of PIS As and TSR violations occurred due to inadequate or erroneous system design information. As a corrective action, BJC assessed the occurrences caused by systems design-safety basis interface problems. Safety systems reviewed included the Molten Salt Reactor Experiment (MSRE) Fluorination System, K-1065 fire alarm system, and the K-25 Radiation Criticality Accident Alarm System. The conclusion was that an inadequate knowledge of system design could result in continuous non-compliance issues relating to nuclear safety. This was especially true with older facilities that lacked current as-built drawings coupled with the loss of 'historical knowledge' as personnel retired or moved on in their careers. Walkdown of systems and the updating of drawings are imperative for nuclear safety compliance. System design integration with safety basis has relevance in the Department of Energy (DOE) complex. This paper presents the BJC Lessons Learned in this area. It will be of benefit to DOE contractors that manage and operate an aging population of nuclear facilities.

  2. The Team-Based Operation of Safety-Critical, Programmable Systems in US Commercial Aviation and the UK Maritime Industries

    E-Print Network [OSTI]

    Johnson, C.W.

    Johnson,C.W. In C.W. Johnson and P. Palanque (eds), Human Error, Safety and Systems Development pp 255-270 Kluwer

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

    E-Print Network [OSTI]

    Hill, Wendell T.

    Research on nuclear energy started at the University of Maryland just after World War II, when and nuclear weapons was followed by controversial accidents and regulation. Today, nuclear power is considered that analyze the risks involved in the use of nuclear energy. Understanding and Using Radiation The ionizing

  4. 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

    2013-10-01T23:59:59.000Z

    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.

  5. Oak Ridge National Laboratory Health and Safety Long-Range Plan: Fiscal years 1989--1995

    SciTech Connect (OSTI)

    Not Available

    1989-06-01T23:59:59.000Z

    The health and safety of its personnel is the first concern of ORNL and its management. The ORNL Health and Safety Program has the responsibility for ensuring the health and safety of all individuals assigned to ORNL activities. This document outlines the principal aspects of the ORNL Health and Safety Long-Range Plan and provides a framework for management use in the future development of the health and safety program. Each section of this document is dedicated to one of the health and safety functions (i.e., health physics, industrial hygiene, occupational medicine, industrial safety, nuclear criticality safety, nuclear facility safety, transportation safety, fire protection, and emergency preparedness). Each section includes functional mission and objectives, program requirements and status, a summary of program needs, and program data and funding summary. Highlights of FY 1988 are included.

  6. Cognitive decision errors and organization vulnerabilities in nuclear power plant safety management: Modeling using the TOGA meta-theory framework

    SciTech Connect (OSTI)

    Cappelli, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Gadomski, A. M. [ECONA, Centro Interuniversitario Elaborazione Cognitiva Sistemi Naturali e Artificiali, via dei Marsi 47, Rome (Italy); Sepiellis, M. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Wronikowska, M. W. [UTFISST, ENEA Casaccia, via Anguillarese 301, Rome (Italy); Poznan School of Social Sciences (Poland)

    2012-07-01T23:59:59.000Z

    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)

  7. Critical analysis of quark-meson coupling models for nuclear matter and finite nuclei

    E-Print Network [OSTI]

    Horst Mueller; Byron K. Jennings

    1998-07-09T23:59:59.000Z

    Three versions of the quark-meson coupling (QMC) model are applied to describe properties of nuclear matter and finite nuclei. The models differ in the treatment of the bag constant and in terms of nonlinear scalar self-interactions. As a consequence opposite predictions for the medium modifications of the internal nucleon structure arise. After calibrating the model parameters at equilibrium nuclear matter density, binding energies, charge radii, single-particle spectra and density distributions of spherical nuclei are analyzed and compared with QHD calculations. For the models which predict a decreasing size of the nucleon in the nuclear environment, unrealistic features of the nuclear shapes arise.

  8. Safety research programs sponsored by Office of Nuclear Regulatory Research. Volume 3. No. 2. Quarterly progress report, April 1-June 30, 1983

    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

    1983-09-01T23:59:59.000Z

    The projects reported are the following: HTGR Safety Evaluation, SSC development, Validation and Application, CRBR balance of plant modeling, thermal-hydraulic reactor safety experiments, LWR plant analyzer development, LWR code assessment and application, thermal reactor code development (RAMONA-3B); stress corrosion cracking of PWR steam generator tubing, bolting failure analysis, probability based load combinations for design of category I structures, mechanical piping benchmark problems; human error data for nuclear power plant safety related events, criteria for human engineering regulatory guides and human factors in nuclear power plant safeguards.

  9. Safety research programs sponsored by Office of Nuclear Regulatory Research. Quarterly progress report, July 1-September 30, 1983. Volume 3, No. 3

    SciTech Connect (OSTI)

    Weiss, A J [comp.

    1984-01-01T23:59:59.000Z

    The projects reported are the following: HTGR Safety Evaluation, SSC Development, Validation and Application, CRBR Balance of Plant Modeling, Thermal-Hydraulic Reactor Safety Experiments, LWR Plant Analyzer Development, LWR Code Assessment and Application, Thermal Reactor Code Development (RAMONA-3B); Stress Corrosion Cracking of PWR Steam Generator Tubing, Bolting Failure Analysis, Probability Based Load Combinations for Design of Category I Structures, Mechanical Piping Benchmark Problems; Human Error Data for Nuclear Power Plant Safety-Related Events, and Human Factors in Nuclear Power Plant Safeguards.

  10. 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

    1983-06-01T23:59:59.000Z

    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.

  11. Nuclear reactor safety. Quarterly progress report, October 1-December 31, 1979

    SciTech Connect (OSTI)

    Jackson, J.F.; Stevenson, M.G. (comps.)

    1980-05-01T23:59:59.000Z

    Progress is reported in the following areas: LWRs, LMFBRs, HTGRs, GCFRs, and safety analysis of the TMI-2 severe overcooling accident. (DLC)

  12. Regulatory analysis for the resolution of Generic Safety Issue 29: Bolting degradation or failure in nuclear power plants

    SciTech Connect (OSTI)

    Chang, T.Y.

    1991-09-01T23:59:59.000Z

    Generic Safety Issue (GSI)-29 deals with staff concerns about public risk due to degradation or failure of safety-related bolting in nuclear power plants. The issue was initiated in November 1982. Value-impact studies of a mandatory program on safety-related bolting for operating plants were inconclusive: therefore, additional regulatory requirements for operating plants could not be justified in accordance with provisions of 10 CFR 50.109. In addition, based on operating experience with bolting in both nuclear and conventional power plants, the actions already taken through bulletins, generic letters, and information notices, and the industry-proposed actions, the staff concluded that a sufficient technical basis exists for the resolution of GSI-29. The staff further concluded that leakage of bolted pressure joints is possible but catastrophic failure of a reactor coolant pressure boundary joint that will lead to significant accident sequences is highly unlikely. For future plants, it was concluded that a new Standard Review Plant section should be developed to codify existing bolting requirements and industry-developed initiatives. 9 refs., 1 tab.

  13. Space reactor safety, 1985--1995 lessons learned

    SciTech Connect (OSTI)

    Marshall, A.C.

    1995-12-31T23:59:59.000Z

    Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

  14. Space reactor safety, 1985{endash}1995 lessons learned

    SciTech Connect (OSTI)

    Marshall, A.C. [Sandia National Laboratories, P.O. Box 5800, Albuquerque, New Mexico 87185 (United States)

    1996-03-01T23:59:59.000Z

    Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatics, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration. {copyright} {ital 1996 American Institute of Physics.}

  15. Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository--BMT1 of the DECOVALEX III project. Part 2: Effects of THM

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste, Germany (7) Canadian Nuclear Safety Commission (CNSC), Ottawa, Canada (8) Lawrence Berkeley National Laboratory (LBNL), Berkeley, USA (9) INERIS-LAEGO, Ecole des Mines de Nancy, Nancy, France (10) Japan Nuclear

  16. Criticality concerns in cleaning large uranium hexafluoride cylinders

    SciTech Connect (OSTI)

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

    1995-06-01T23:59:59.000Z

    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. Safety evaluation report related to the operation of Watts Bar Nuclear Plant, Units 1 and 2 (Docket Nos. 50-390 and 50-391). Supplement No. 14

    SciTech Connect (OSTI)

    Tam, P.S.

    1994-12-01T23:59:59.000Z

    Supplement No. 14 to the Safety Evaluation Report for the application filed by the Tennessee Valley Authority for license to operate Watts Bar Nuclear Plant, Units 1 and 2, Docket Nos. 50-390 and 50-391, located in Rhea County, Tennessee, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. The purpose of this supplement is to update the Safety Evaluation with additional information submitted by the applicant since Supplement No. 13 was issued, and matters that the staff had under review when Supplement No. 13 was issued.

  18. Rewriting the standard on the functional requirements for computers used in safety systems of nuclear power plants

    SciTech Connect (OSTI)

    Matras, J.R. [Science Applications International Corp., Las Vegas, NV (United States)

    1991-07-01T23:59:59.000Z

    Because of the rapid development of state-of-the-art computer technology, a rewrite of ANSI/IEEE-ANS-7-4.3.2-1982, {open_quotes}Application Criteria for Digital Computer Systems of Nuclear Power Generating Stations,{close_quotes} was required. This article outlines the thrust of this rewrite, which is nearing the balloting process, and identifies standards and guidelines to be used in the development of a highly reliable to be used in the development of a highly reliable computer system. The rewrite activity has been in process approximately 2 yr and is a cooperative project of the American Nuclear Society (ANS) Nuclear Power Plant Standards Committee (NUPPSCO) and the Institute of Electrical and Electronics Engineers (IEEE) Nuclear Power Engineering Committee (NPEC). Because computer technology has progressed significantly since ANSI/IEEE-ANS-7-4.3.2-1982 was issued, the rewrite was a very interesting challenge to the work group. The primary difference between the 1982 version and the rewrite is that the 1982 version addressed the quality assurance aspects of the Quality criteria, which included the integration of hardware and software and subsequent verification and validation, whereas the rewrite, being a product of IEEE Std 603-1991, {open_quotes}IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations,{close_quotes} was written to establish the minimum requirements for computer systems (hardware, software, and interfaces) used in safety systems. This article presents an introduction to the scope of the rewrite, provides a brief comment on how the work group addressed the scope, and follows with details on how the work group addressed the scope and rewrite direction. 1 fig., 1 tab.

  19. 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

    2010-11-01T23:59:59.000Z

    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.

  20. Criticality Safety Analysis on the Mixed Be, Nat-U, and C (Graphite) Reflectors in 55-Gallon Waste Drums and Their Equivalents for HWM Applications

    SciTech Connect (OSTI)

    Chou, P

    2011-12-14T23:59:59.000Z

    The objective of this analysis is to develop and establish the technical basis on the criticality safety controls for the storage of mixed beryllium (Be), natural uranium (Nat-U), and carbon (C)/graphite reflectors in 55-gallon waste containers and/or their equivalents in Hazardous Waste Management (HWM) facilities. Based on the criticality safety limits and controls outlined in Section 3.0, the operations involving the use of mixed-reflector drums satisfy the double-contingency principle as required by DOE Order 420.1 and are therefore criticality safe. The mixed-reflector mass limit is 120 grams for each 55-gallon drum or its equivalent. a reflector waiver of 50 grams is allowed for Be, Nat-U, or C/graphite combined. The waived reflectors may be excluded from the reflector mass calculations when determining if a drum is compliant. The mixed-reflector drums are allowed to mix with the typical 55-gallon one-reflector drums with a Pu mass limit of 120 grams. The fissile mass limit for the mixed-reflector container is 65 grams of Pu equivalent each. The corresponding reflector mass limits are 300 grams of Be, and/or 100 kilograms of Nat-U, and/or 110 kilograms of C/graphite for each container. All other unaffected control parameters for the one-reflector containers remain in effect for the mixed-reflector drums. For instance, Superior moderators, such as TrimSol, Superla white mineral oil No. 9, paraffin, and polyethylene, are allowed in unlimited quantities. Hydrogenous materials with a hydrogen density greater than 0.133 gram/cc are not allowed. Also, an isolation separation of no less than 76.2 cm (30-inch) is required between a mixed array and any other array. Waste containers in the action of being transported are exempted from this 76.2-cm (30-inch) separation requirement. All deviations from the CS controls and mass limits listed in Section 3.0 will require individual criticality safety analyses on a case-by-case basis for each of them to confirm their criticality safety prior to their deployment and implementation.