Sample records for nrc nuclear regulatory

  1. NRC (Nuclear Regulatory Commission) perspective of software QA (quality assurance) in the nuclear history

    SciTech Connect (OSTI)

    Weiss, S.H.

    1988-01-01T23:59:59.000Z

    Computer technology has been a part of the nuclear industry since its inception. However, it is only recently that computers have been integrated into reactor operations. During the early history of commercial nuclear power in the United States, the US Nuclear Regulatory Commission (NRC) discouraged the use of digital computers for real-time control and monitoring of nuclear power plant operation. At the time, this position was justified since software engineering was in its infancy, and horror stories on computer crashes were plentiful. Since the advent of microprocessors and inexpensive computer memories, significant advances have been made in fault-tolerant computer architecture that have resulted in highly reliable, durable computer systems. The NRC's requirement for safety parameter display system (SPDS) stemmed form the results of studies and investigations conducted on the Three Mile Island Unit 2 (TMI-2) accident. An NRC contractor has prepared a handbook of software QA techniques applicable to the nuclear industry, published as NUREG/CR-4640 in August 1987. Currently, the NRC is considering development of an inspection program covering software QA. Future efforts may address verification and validation as applied to expert systems and artificial intelligence programs.

  2. Comparison of the NRC and the IAEA regulatory documents in the area of nuclear fuel systems

    SciTech Connect (OSTI)

    El-Adham, K.; Shinaishin, M.A.

    1991-04-01T23:59:59.000Z

    A main objective of this work was to identify the safety requirements in the area of fuel system design and performance from both the International Atomic Energy Agency (IAEA) and U.S. Nuclear Regulatory Commission (NRC) points of view. The study covered requirements during normal plant operation as well as during accident conditions. This study revealed that, although none of the factors to be considered for fuel safety were neglected in the IAEA regulatory documents, these documents are not complete in themselves, particularly because they lack quantitative guidelines and specific industrial standards. Although generality makes the IAEA requirements adaptable to many countries, on the other hand, it makes their applicability constrained by the availability of highly qualified and experienced personnel who can translate the qualitative requirements given in these documents into actual engineering solutions. 20 refs.

  3. Enhancement of NRC station blackout requirements for nuclear power plants

    SciTech Connect (OSTI)

    McConnell, M. W. [United States Nuclear Regulatory Commission, Mail Stop: 012-H2, Washington, DC 20555 (United States)

    2012-07-01T23:59:59.000Z

    The U.S. Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) in response to Commission direction to conduct a systematic and methodical review of NRC processes and regulations to determine whether the agency should make additional improvements to its regulatory system and to make recommendations to the Commission for its policy direction, in light of the accident at the Fukushima Dai-ichi Nuclear Power Plant. The NTTF's review resulted in a set of recommendations that took a balanced approach to defense-in-depth as applied to low-likelihood, high-consequence events such as prolonged station blackout (SBO) resulting from severe natural phenomena. Part 50, Section 63, of Title 10 of the Code of Federal Regulations (CFR), 'Loss of All Alternating Current Power,' currently requires that each nuclear power plant must be able to cool the reactor core and maintain containment integrity for a specified duration of an SBO. The SBO duration and mitigation strategy for each nuclear power plant is site specific and is based on the robustness of the local transmission system and the transmission system operator's capability to restore offsite power to the nuclear power plant. With regard to SBO, the NTTF recommended that the NRC strengthen SBO mitigation capability at all operating and new reactors for design-basis and beyond-design-basis external events. The NTTF also recommended strengthening emergency preparedness for prolonged SBO and multi-unit events. These recommendations, taken together, are intended to clarify and strengthen US nuclear reactor safety regarding protection against and mitigation of the consequences of natural disasters and emergency preparedness during SBO. The focus of this paper is on the existing SBO requirements and NRC initiatives to strengthen SBO capability at all operating and new reactors to address prolonged SBO stemming from design-basis and beyond-design-basis external events. The NRC initiatives are intended to enhance core and spent fuel pool cooling, reactor coolant system integrity, and containment integrity. (authors)

  4. Nuclear Regulatory Commission Information Digest 1992 edition. Volume 4

    SciTech Connect (OSTI)

    Olive, K [ed.] [ed.

    1992-03-01T23:59:59.000Z

    The Nuclear Regulatory Commission Information Digest provides a summary of information about the US Nuclear Regulatory Commission (NRC), NRC's regulatory responsibilities, the activities NRC licenses, and general information on domestic and worldwide nuclear energy. This digest is a compilation of nuclear- and NRC-related data and is designed to provide a quick reference to major facts about the agency and industry it regulates. In general, the data cover 1975 through 1991, with exceptions noted. Information on generating capacity and average capacity factor for operating US commercial nuclear power reactors is obtained from monthly operating reports that are submitted directly to the NRC by the licensee. This information is reviewed by the NRC for consistency only and no independent validation and/or verification is performed.

  5. Nuclear Regulatory Commission issuances

    SciTech Connect (OSTI)

    NONE

    1996-03-01T23:59:59.000Z

    This document is the March 1996 listing of NRC issuances. Included are: (1) NRC orders granting Cleveland Electric Illuminating Company`s petition for review of the ASLB order LBP-95-17, (2) NRC orders relating to the potential disqualification of two commissioners in the matter of the decommissioning of Yankee Nuclear Power Station, (3) ASLB orders pertaining to the Oncology Services Corporation, (4) ASLB orders pertaining to the Radiation Oncology Center, (5) ASLB orders pertaining to the Yankee Nuclear Power Station, and (6) Director`s decision pertaining to the Yankee Nuclear Power Station.

  6. Nuclear Regulatory Commission Information Digest 1994 edition. Volume 6

    SciTech Connect (OSTI)

    Stadler, L.

    1994-03-01T23:59:59.000Z

    The Nuclear Regulatory Commission Information Digest (digest) provides a summary of information about the US Nuclear Regulatory Commission (NRC), NRCs regulatory responsibility the activities NRC licenses, and general information on domestic and worldwide nuclear energy. The digest, published annually, is a compilation of nuclear-and NRC-related data and is designed to provide a quick reference to major facts about the agency and the industry it regulates. In general, the data cover 1975 through 1993, with exceptions noted. Information on generating capacity and average capacity factor for operating US commercial nuclear power reactors is obtained from monthly operating reports that are submitted directly to the NRC by the information is reviewed by the NRC for consistency only and no independent validation and/or verification is performed.

  7. Recommendations for NRC policy on shift scheduling and overtime at nuclear power plants

    SciTech Connect (OSTI)

    Lewis, P.M.

    1985-07-01T23:59:59.000Z

    This report contains the Pacific Northwest Laboratory's (PNL's) recommendations to the US Nuclear Regulatory Commission (NRC) for an NRC policy on shift scheduling and hours of work (including overtime) for control room operators and other safety-related personnel in nuclear power plants. First, it is recommended that NRC make three additions to its present policy on overtime: (1) limit personnel to 112 hours of work in a 14-day period, 192 hours in 28 days, and 2260 hours in one year; exceeding these limits would require plant manager approval; (2) add a requirement that licensees obtain approval from NRC if plant personnel are expected to exceed 72 hours of work in a 7-day period, 132 hours in 14 days, 228 hours in 28 days, and 2300 hours in one year; and (3) make the policy a requirement, rather than a nonbinding recommendation. Second, it is recommended that licensees be required to obtain NRC approval to adopt a routine 12-hour/day shift schedule. Third, it is recommended that NRC add several nonbinding recommendations concerning routine 8-hour/day schedules. Finally, because additional data can strengthen the basis for future NRC policy on overtime, five methods are suggested for collecting data on overtime and its effects. 44 refs., 10 tabs.

  8. Nuclear regulatory legislation, 104th Congress, Volume 1, No. 4

    SciTech Connect (OSTI)

    NONE

    1997-12-01T23:59:59.000Z

    This document is the first of two volumes compiling statutes and material pertaining to nuclear regulatory legislation through the 104th Congress, 2nd Session. It is intended for use as a U.S. Nuclear Regulatory Commission (NRC) internal resource document. Legislative information reproduced in this document includes portions of the Atomic Energy Act, Energy Reorganization Act, Low-Level Radioactive Waste Policy Amendments Act, and Nuclear Waste Policy Act. Other information included in this volume pertains to NRC user fees, NRC authorizations, the Inspector General Act, and the Administrative Procedure Act.

  9. Accident Conditions versus Regulatory Test for NRC-Approved UF6 Packages

    SciTech Connect (OSTI)

    MILLS, G. SCOTT; AMMERMAN, DOUGLAS J.; LOPEZ, CARLOS

    2003-01-01T23:59:59.000Z

    The Nuclear Regulatory Commission (NRC) approves new package designs for shipping fissile quantities of UF{sub 6}. Currently there are three packages approved by the NRC for domestic shipments of fissile quantities of UF{sub 6}: NCI-21PF-1; UX-30; and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR Part 71. The primary objective of this project was to relate the conditions experienced by these packages in the tests described in 10 CFR Part 71 to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR Part 71 tests was achieved by means of computer modeling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from test and other fire scenarios. In addition, the likelihood of encountering bodies of water or sufficient rainfall to cause complete or partial immersion during transport over representative truck routes was assessed. Modeled effects, and their associated probabilities, were combined with existing event-tree data, plus accident rates and other characteristics gathered from representative routes, to derive generalized probabilities of encountering accident conditions comparable to the 10 CFR Part 71 conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents, i.e. the likelihood of UF{sub 6} being dispersed as a result of accident impact or fire is small. Moreover, given that an accident has occurred, exposure to water by fire-fighting, heavy rain or submersion in a body of water is even less probable by factors ranging from 0.5 to 8E-6.

  10. Proceedings of the 21st DOE/NRC nuclear air cleaning conference; Volume 2, Sessions 9--16

    SciTech Connect (OSTI)

    First, M.W. [ed.] [Harvard Univ., Boston, MA (United States). Harvard Air Cleaning Lab.

    1991-02-01T23:59:59.000Z

    The 21st meeting of the Department of Energy/Nuclear Regulatory Commission (DOE/NRC) Nuclear Air Cleaning Conference was held in San Diego, CA on August 13--16, 1990. The proceedings have been published as a two volume set. Volume 2 contains sessions covering adsorbents, nuclear codes and standards, modelling, filters, safety, containment venting and a review of nuclear air cleaning programs around the world. Also included is the list of attendees and an index of authors and speakers. (MHB)

  11. NRC Technical Research Program to Evaluate Extended Storage and Transportation of Spent Nuclear Fuel - 12547

    SciTech Connect (OSTI)

    Einziger, R.E.; Compton, K.; Gordon, M.; Ahn, T.; Gonzales, H. [United States Nuclear Regulatory Commission, Rockville, Maryland 20852 (United States); Pan, Y. [Center for Nuclear Waste Regulatory Analyses, San Antonio, TX 78238 (United States)

    2012-07-01T23:59:59.000Z

    Any new direction proposed for the back-end of spent nuclear fuel (SNF) cycle will require storage of SNF beyond the current licensing periods. The Nuclear Regulatory Commission (NRC) has established a technical research program to determine if any changes in the 10 CFR part 71, and 72 requirements, and associated guidance might be necessary to regulate the safety of anticipated extended storage, and subsequent transport of SNF. This three part program of: 1) analysis of knowledge gaps in the potential degradation of materials, 2) short-term research and modeling, and 3) long-term demonstration of systems, will allow the NRC to make informed regulatory changes, and determine when and if additional monitoring and inspection of the systems is necessary. The NRC has started a research program to obtain data necessary to determine if the current regulatory guidance is sufficient if interim dry storage has to be extended beyond the currently approved licensing periods. The three-phased approach consists of: - the identification and prioritization of potential degradation of the components related to the safe operation of a dry cask storage system, - short-term research to determine if the initial analysis was correct, and - a long-term prototypic demonstration project to confirm the models and results obtained in the short-term research. The gap analysis has identified issues with the SCC of the stainless steel canisters, and SNF behavior. Issues impacting the SNF and canister internal performance such as high and low temperature distributions, and drying have also been identified. Research to evaluate these issues is underway. Evaluations have been conducted to determine the relative values that various types of long-term demonstration projects might provide. These projects or follow-on work is expected to continue over the next five years. (authors)

  12. Nuclear Regulatory legislation: 103d Congress. Volume 1, No. 3

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    This document is a compilation of nuclear regulatory legislation and other relevant material through the 103d Congress, 2d Session. This compilation has been prepared for use as a resource document, which the NRC intends to update at the end of every Congress. The contents of NUREG-0980 include the Atomic Energy Act of 1954, as amended; Energy Reorganization Act of 1974, as amended, Uranium Mill Tailings Radiation Control Act of 1978; Low-Level Radioactive Waste Policy Act; Nuclear Waste Policy Act of 1982; and NRC Authorization and Appropriations Acts. Other materials included are statutes and treaties on export licensing, nuclear non-proliferation, and environmental protection.

  13. Nuclear Regulatory legislation: 103d Congress. Volume 2, No. 3

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    This document is a compilation of nuclear regulatory legislation and other relevant material through the 103d Congress, 2d Session. This compilation has been prepared for use as a resource document, which the NRC intends to update at the end of every Congress. The contents of NUREG-0980 include the Atomic Energy Act of 1954, as amended; Energy Reorganization Act of 1974, as amended, Uranium Mill Tailings Radiation Control Act of 1978; Low-Level Radioactive Waste Policy Act; Nuclear Waste Policy Act of 1982; and NRC Authorization and Appropriations Acts. Other materials included are statutes and treaties on export licensing, nuclear non-proliferation, and environmental protection.

  14. Nuclear regulatory legislation: 102d Congress. Volume 1, No. 2

    SciTech Connect (OSTI)

    Not Available

    1993-10-01T23:59:59.000Z

    This document is a compilation of nuclear regulatory legislation and other relevant material through the 102d Congress, 2d Session. This compilation has been prepared for use as a resource document, which the NRC intends to update at the end of every Congress. The contents of NUREG-0980 include: The Atomic Energy Act of 1954, as amended; Energy Reorganization Act of 1974, as amended, Uranium Mill Tailings Radiation Control Act of 1978; Low-Level Radioactive Waste Policy Act; Nuclear Waste Policy Act of 1982; and NRC Authorization and Appropriations Acts. Other materials included are statutes and treaties on export licensing, nuclear non-proliferation, and environmental protection.

  15. Nuclear regulatory legislation, 102d Congress. Volume 2, No. 2

    SciTech Connect (OSTI)

    Not Available

    1993-10-01T23:59:59.000Z

    This document is a compilation of nuclear regulatory legislation and other relevant material through the 102d Congress, 2d Session. This compilation has been prepared for use as a resource document, which the NRC intends to update at the end of every Congress. The contents of NUREG-0980 include The Atomic Energy Act of 1954, as amended; Energy Reorganization Act of 1974, as amended, Uranium Mill Tailings Radiation Control Act of 1978; Low-Level Radioactive Waste Policy Act; Nuclear Waste Policy Act of 1982; and NRC Authorization and Appropriations Acts. Other materials included are statutes and treaties on export licensing, nuclear non-proliferation, and environmental protection.

  16. Energy Department Issues Request For Proposal for Nuclear Regulatory...

    Office of Environmental Management (EM)

    Facilities procurement. The NRC Licensed Facilities contract is for managing Spent Nuclear Fuel (SNF) storage facilities and licenses under NRC regulations. The scope...

  17. Nuclear regulatory legislation, 104th Congress. Volume 2, No. 4

    SciTech Connect (OSTI)

    NONE

    1997-12-01T23:59:59.000Z

    This document is the second of two volumes compiling statutes and material pertaining to nuclear regulatory legislation through the 104th Congress, 2nd Session. It is intended for use as a U.S. Nuclear Regulatory Commission (NRC) internal resource document. Legislative information reproduced in this document includes portions of the Paperwork Reduction Act, various acts pertaining to low-level radioactive waste, the Clean Air Act, the Federal Water Pollution Control Act, the National Environmental Policy Act, the Hazardous Materials Transportation Act, the West Valley Demonstration Project Act, Nuclear Non-Proliferation and Export Licensing Statutes, and selected treaties, agreements, and executive orders. Other information provided pertains to Commissioner tenure, NRC appropriations, the Chief Financial Officers Act, information technology management reform, and Federal civil penalties.

  18. NRC antitrust licensing actions, 1978--1996

    SciTech Connect (OSTI)

    Mayer, S.J.; Simpson, J.J.

    1997-09-01T23:59:59.000Z

    NUREG-0447, Antitrust Review of Nuclear Power Plants, was published in May 1978 and includes a compilation and discussion of U.S. Nuclear Regulatory Commission (NRC) proceedings and activity involving the NRC`s competitive review program through February 1978, NUREG-0447 is an update of an earlier discussion of the NRC`s antitrust review of nuclear power plants, NR-AIG-001, The US Nuclear Regulatory Commission`s Antitrust Review of Nuclear Power Plants: The Conditioning of Licenses, which reviewed the Commission`s antitrust review function from its inception in December 1970 through April 1976. This report summarizes the support provided to NRC staff in updating the compilation of the NRC`s antitrust licensing review activities for commercial nuclear power plants that have occurred since February 1978. 4 refs., 4 tabs.

  19. Preservation and Implementation of Decommissioning Lessons Learned in the United States Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Rodriguez, Rafael L. [United States Nuclear Regulatory Commission, Office of Federal and State Materials and Environmental Management Programs, Washington, DC 20555 (United States)

    2008-01-15T23:59:59.000Z

    Over the past several years, the United States Nuclear Regulatory Commission (NRC) has actively worked to capture and preserve lessons learned from the decommissioning of nuclear facilities. More recently, NRC has involved industry groups, the Organization of Agreement States (OAS), and the Department of Energy (DOE) in the effort to develop approaches to capture, preserve and disseminate decommissioning lessons learned. This paper discusses the accomplishments of the working group, some lessons learned by the NRC in the recent past, and how NRC will incorporate these lessons learned into its regulatory framework. This should help ensure that the design and operation of current and future nuclear facilities will result in less environmental impact and more efficient decommissioning. In summary, the NRC will continue capturing today's experience in decommissioning so that future facilities can take advantage of lessons learned from today's decommissioning projects. NRC, both individually and collectively with industry groups, OAS, and DOE, is aggressively working on the preservation and implementation of decommissioning lessons learned. The joint effort has helped to ensure the lessons from the whole spectrum of decommissioning facilities (i.e., reactor, fuel cycle, and material facilities) are better understood, thus maximizing the amount of knowledge and best practices obtained from decommissioning activities. Anticipated regulatory activities at the NRC will make sure that the knowledge gained from today's decommissioning projects is preserved and implemented to benefit the nuclear facilities that will decommission in the future.

  20. UNITED STATES NUCLEAR REGULATORY COMMISSION

    E-Print Network [OSTI]

    UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS ON RADIATION THERAPY OVEREXPOSURES IN PANAMA Addressees All medical licensees. Purpose The U.S. Nuclear persons in your institution who are involved with radiation therapy should review this notice

  1. Twenty-third DOE/NRC nuclear air cleaning and treatment conference

    SciTech Connect (OSTI)

    Bellamy, R.R.; Hayes, J.J.; First, M.W.

    1995-01-01T23:59:59.000Z

    The Twenty-Third Department of Energy/Nuclear Regulatory Commission (DOE/NRC) nuclear Air-Cleaning and Treatment Conference was held July 25-28, 1994, in Buffalo, New York. The conference was also sponsored by the Harvard Air-Cleaning Laboratory and the Internation Society of Nuclear Air Treatment Technologies, a nonprofit organization founded to promote technology transfer in the nuclear air-cleaning and treatment area. A total of 192 air-cleaning specialists attended the conference. The United States and 11 foreign countries were represented. The specialists are affiliated with all aspects of the nuclear industry, including government agencies, educational institutions, utilities, architect-engineers, equipment suppliers, and consultants. The high level of international interests is evident from the 40% of papers sponsored by foreign interests. More than 20% of the attendees as well as several members of the Program Committee were from outside the United States. Major topics discussed at this conference included nuclear air-cleaning codes and standards, waste disposal, particulate filter developments (including testing and performance under stress and after aging), sampling and monitoring of process and effluent streams, off-gasses from fuel reprocessing, adsorbents and adsorption, accident control and analysis, and revised source terms for power-plant accidents. A highlight of the conference concerned operations a at the DOE facility at West Valley, New York, where construction is under way to solidify radioactive waste. A recurrent theme throughout the sessions was that, in spite of the large number of guidance documents available in the form of regulations, codes, standards, and directives, multiple difficulties arise when all are invoked simultaneously. Gas processing needs, rather than controls for civilian power plants, will provide the principal challenge during the next decade for the air-cleaning specialists of the world. 15 refs.

  2. 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...

  3. EPRI/NRC-RES fire PRA guide for nuclear power facilities. Volume 1, summary and overview.

    SciTech Connect (OSTI)

    Not Available

    2004-09-01T23:59:59.000Z

    This report documents state-of-the-art methods, tools, and data for the conduct of a fire Probabilistic Risk Assessment (PRA) for a commercial nuclear power plant (NPP) application. The methods have been developed under the Fire Risk Re-quantification Study. This study was conducted as a joint activity between EPRI and the U. S. NRC Office of Nuclear Regulatory Research (RES) under the terms of an EPRI/RES Memorandum of Understanding [RS.1] and an accompanying Fire Research Addendum [RS.2]. Industry participants supported demonstration analyses and provided peer review of this methodology. The documented methods are intended to support future applications of Fire PRA, including risk-informed regulatory applications. The documented method reflects state-of-the-art fire risk analysis approaches. The primary objective of the Fire Risk Study was to consolidate recent research and development activities into a single state-of-the-art fire PRA analysis methodology. Methodological issues raised in past fire risk analyses, including the Individual Plant Examination of External Events (IPEEE) fire analyses, have been addressed to the extent allowed by the current state-of-the-art and the overall project scope. Methodological debates were resolved through a consensus process between experts representing both EPRI and RES. The consensus process included a provision whereby each major party (EPRI and RES) could maintain differing technical positions if consensus could not be reached. No cases were encountered where this provision was invoked. While the primary objective of the project was to consolidate existing state-of-the-art methods, in many areas, the newly documented methods represent a significant advancement over previously documented methods. In several areas, this project has, in fact, developed new methods and approaches. Such advances typically relate to areas of past methodological debate.

  4. 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

  5. Regulatory Endorsement Activities for ASME Nuclear Codes and Standards

    SciTech Connect (OSTI)

    West, Raymond A. [Dominion Nuclear Connecticut, Inc. - Dominion Generation (United States)

    2006-07-01T23:59:59.000Z

    The ASME Board on Nuclear Codes and Standards (BNCS) has formed a Task Group on Regulatory Endorsement (TG-RE) that is currently in discussions with the United States Nuclear Regulatory Commission (NRC) to look at suggestions and recommendations that can be used to help with the endorsement of new and revised ASME Nuclear Codes and Standards (NC and S). With the coming of new reactors in the USA in the very near future we need to look at both the regulations and all the ASME NC and S to determine where we need to make changes to support these new plants. At the same time it is important that we maintain our operating plants while addressing ageing management needs of our existing reactors. This is going to take new thinking, time, resources, and money. For all this to take place the regulations and requirements that we use must be clear concise and necessary for safety and to that end both the NRC and ASME are working together to make this happen. Because of the influence that the USA has in the world in dealing with these issues, this paper is written to inform the international nuclear engineering community about the issues and what actions are being addressed under this effort. (author)

  6. Microsoft PowerPoint - NRC Nuclear Export Controls Implementing...

    National Nuclear Security Administration (NNSA)

    international treaties, agreements, and other obligations * Apply to any person who exports or imports nuclear equipment and material with few exceptions * Authorize export...

  7. 1996 NRC annual report. Volume 13

    SciTech Connect (OSTI)

    NONE

    1997-10-01T23:59:59.000Z

    This 22nd annual report of the US Nuclear Regulatory Commission (NRC) describes accomplishments, activities, and plans made during Fiscal Year 1996 (FH 1996)--October 1, 1995, through September 30, 1996. Significant activities that occurred early in FY 1997 are also described, particularly changes in the Commission and organization of the NRC. The mission of the NRC is to ensure that civilian uses of nuclear materials in the US are carried out with adequate protection of public health and safety, the environment, and national security. These uses include the operation of nuclear power plants and fuel cycle plants and medical, industrial, and research applications. Additionally, the NRC contributes to combating the proliferation of nuclear weapons material worldwide. The NRC licenses and regulates commercial nuclear reactor operations and research reactors and other activities involving the possession and use of nuclear materials and wastes. It also protects nuclear materials used in operation and facilities from theft or sabotage. To accomplish its statutorily mandated regulatory mission, the NRC issues rules and standards, inspects facilities and operations, and issues any required enforcement actions.

  8. Mr. John Kinneman, Chief Nuclear Materfals Branch Nuclear Regulatory...

    Office of Legacy Management (LM)

    111989 Mr. John Kinneman, Chief Nuclear Materfals Branch Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia. Pennsylvania 19406 Dear Mr. Kinneman: -;' .-. 'W...

  9. Energy Praises the Nuclear Regulatory Commission Approval of...

    Energy Savers [EERE]

    Energy Praises the Nuclear Regulatory Commission Approval of the First United States Nuclear Plant Site in Over 30 Years Energy Praises the Nuclear Regulatory Commission Approval...

  10. Department of Energy and Nuclear Regulatory Commission Increase...

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

    Nuclear Regulatory Commission Increase Cooperation to Advance Global Nuclear Energy Partnership Department of Energy and Nuclear Regulatory Commission Increase Cooperation to...

  11. 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

  12. Nuclear Crisis Communications: The Plan Worked. A Critique of NRC Communications in the Fukushima Daiichi Reactor Crisis - 12073

    SciTech Connect (OSTI)

    Brenner, Eliot; Harrington, Holly; Schmidt, Rebecca [U.S. Nuclear Regulatory Commission, Rockville, MD 20852 (United States)

    2012-07-01T23:59:59.000Z

    'Call the AV-Photo folks and get someone in here to shoot b-roll. We'll never be able to accommodate the network cameras and the only way I can get this to the media is to produce it ourselves'. Eliot Brenner, Director NRC Office of Public Affairs, March 12, 2011. For the past four years we have been speaking to audiences at Waste Management about communications issues. Last year, though we were kept from attending because of the federal budget crisis, our surrogates described to you the lessons the nuclear industry should draw from the BP Gulf oil spill crisis. Those remarks were delivered 11 days before the Fukushima Daiichi tragedy became the nuclear landmark of a generation - an industry changing event with worldwide ramifications, both in science and regulation and in communications. Eliot Brenner cut his teeth on crisis communication in the aviation industry where tragedy unfolds rapidly. He has been a speech-writer to three cabinet secretaries, spokesman for the Federal Aviation Administration and now spokesman for the Nuclear Regulatory Commission since 2004. Holly Harrington manages the NRC crisis response program and has 26 years federal public affairs experience, including eight years at the Federal Emergency Management Agency. Her crisis experience includes the 1989 Loma Prieta earthquake, numerous hurricanes and floods, Sept 11, and, now Fukushima Daiichi. Rebecca Schmidt is a veteran government relations professional whose decades in Washington include service with the House Armed Services Committee, the House Budget Committee and the Secretary of Defense. Collectively, the Offices of Public Affairs and Congressional Affairs conducted the largest outreach for the agency since Three Mile Island. We worked with the basic rule, described to Waste Management last year just 11 days before Fukushima - communicate early, often and clearly. The response - while not without its problems and lessons - went as smoothly as a chaotic event like Fukushima could go. That was due in large measure to the fact that the NRC has a well-tested system of responding to nuclear emergencies, and we followed our plan. (authors)

  13. US Nuclear Regulatory Commission Input to DOE Request for Information...

    Energy Savers [EERE]

    US Nuclear Regulatory Commission Input to DOE Request for Information Smart Grid Implementation Input US Nuclear Regulatory Commission Input to DOE Request for Information Smart...

  14. Managing aging in nuclear power plants: Insights from NRC maintenance team inspection reports

    SciTech Connect (OSTI)

    Fresco, A.; Subudhi, M.; Gunther, W.; Grove, E.; Taylor, J. [Brookhaven National Lab., Upton, NY (United States)

    1993-12-01T23:59:59.000Z

    A plant`s maintenance program is the principal vehicle through which age-related degradation is managed. From 1988 to 1991, the NRC evaluated the maintenance program of every nuclear power plant in the United States. Forty-four out of a total of 67 of the reports issued on these in-depth team inspections were reviewed for insights into the strengths and weaknesses of the programs as related to the need to understand and manage the effects of aging on nuclear plant systems, structures, and components. Relevant information was extracted from these inspection reports and sorted into several categories, including Specific Aging Insights, Preventive Maintenance, Predictive Maintenance and Condition Monitoring, Post Maintenance Testing, Failure Trending, Root Cause Analysis and Usage of Probabilistic Risk Assessment in the Maintenance Process. Specific examples of inspection and monitoring techniques successfully used by utilities to detect degradation due to aging have been identified. The information also was sorted according to systems and components, including: Auxiliary Feedwater, Main Feedwater, High Pressure Injection for both BWRs and PWRs, Service Water, Instrument Air, and Emergency Diesel Generator Air Start Systems, and Emergency Diesel Generators Air Start Systems, emergency diesel generators, electrical components such as switchgear, breakers, relays, and motor control centers, motor operated valves and check valves. This information was compared to insights gained from the Nuclear Plant Aging Research (NPAR) Program. Attributes of plant maintenance programs where the NRC inspectors felt that improvement was needed to properly address the aging issue also are discussed.

  15. Nuclear Regulatory Commission Handling of Beyond Design Basis Events for Nuclear Power Reactors

    Broader source: Energy.gov [DOE]

    Presenter: Bill Reckley, Chief, Policy and Support Branch, Japan Lessons-Learned Project Directorate, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission US Nuclear Regulatory Commission

  16. Regulatory and technical reports: (Abstract index journal). Compilation for first quarter 1997, January--March

    SciTech Connect (OSTI)

    Sheehan, M.A.

    1997-06-01T23:59:59.000Z

    This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. This compilation is published quarterly and cummulated annually. Reports consist of staff-originated reports, NRC-sponsored conference reports, NRC contractor-prepared reports, and international agreement reports.

  17. 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

  18. Papers on the nuclear regulatory dilemma

    SciTech Connect (OSTI)

    Barkenbus, J.N.; Freeman, S.D.; Weinberg, A.M.

    1985-10-01T23:59:59.000Z

    The four papers contained in this report are titled: (1) From Prescriptive to Performance-Based Regulation of Nuclear Power; (2) Nuclear Regulatory Reform: A Technology-Forcing Approach; (3) Improving the Regulation of Nuclear Power; and (4) Science and Its Limits: The Regulators' Dilemma. These four papers investigate issues relating to the long-term regulation of nuclear energy. They were prepared as part of the Institute for Energy Analysis' project on Nuclear Regulation funded by a grant from the Mellon Foundation and a smaller grant by the MacArthur Foundation. Originally this work was to be supplemented by contributions from the Nuclear Regulatory Commission and from the Department of Energy. These contributions were not forthcoming, and as a result the scope of our investigations was more restricted than we had originally planned.

  19. Review of nuclear power plant offsite power source reliability and related recommended changes to the NRC rules and regulations

    SciTech Connect (OSTI)

    Battle, R.E.; Clark, F.H.; Reddoch, T.W.

    1980-05-01T23:59:59.000Z

    The NRC has stated its concern about the reliability of the offsite power system as the preferred emergency source and about the possible damage to a pressurized water reactor (PWR) that could result from a rapid decay of power grid frequency. ORNL contracted with NRC to provide technical assistance to establish criteria that can be used to evaluate the offsite power system for the licensing of a nuclear power plant. The results of many of the studies for this contract are recommendations to assess and control the power grid during operation. This is because most of the NRC regulations pertaining to the offsite power system are related to the design of the power grid, and we believe that additional emphasis on monitoring the power grid operation will improve the reliability of the nuclear plant offsite power supply. 46 refs., 10 figs.

  20. Nuclear Regulatory Commission issuances. Opinions and decisions of the Nuclear Regulatory Commission with selected orders, July 1, 1994--December 31, 1994. Volume 40, Pages 1--387

    SciTech Connect (OSTI)

    NONE

    1994-12-31T23:59:59.000Z

    The hardbound edition of the Nuclear Regulatory Commission Issuances is a final compilation of the monthly issuances. It includes all of the legal precedents for the agency within a six-month period. Any opinions, decisions, denials, memoranda and orders of the Commission inadvertently omitted from the monthly softbounds and any corrections submitted by the NRC legal staff to the printed softbound issuances are contained in the hardbound edition. Cross references in the text and indexes are to the NRCI page numbers which are the same as the page numbers in this publication. This book covers the following: issuances of the NRC; issuances of the Atomic Safety and Licensing Boards; and issuances of Directors` decisions.

  1. Application of Nuclear Regulatory Commission Regulation Equivalency to Construction of New Nuclear Facilities

    SciTech Connect (OSTI)

    BISHOP, G.E.

    1999-06-02T23:59:59.000Z

    The Spent Nuclear Fuels Project (SNFP) Office of the Department of Energy (DOE), Richland Operations Office, is charged with moving 2.100 metric tons of spent nuclear fuel elements left over from plutonium production into semi-permanent storage at DOE'S Hanford site in Washington state. In anticipation of eventual NRC regulation, the DOE decided to impose NRC requirements on new SNFP facility design and construction, specifically for the Cold Vacuum Drying Facility (CVDF) and the Canister Storage Building (CSB). The SNFP implemented this policy of ''NRC equivalency'' with the goal of achieving a level of nuclear safety equivalent to that of NRC-licensed fuel processing facilities. Appropriate features of the NRC licensing process were adopted. However, the SNFP maintained applicable DOE requirements in tandem with the NRC regulations. Project work is continuing, with the first fuel movement scheduled for November, 2000.

  2. Status of the NRC Decommissioning Program

    SciTech Connect (OSTI)

    Orlando, D. A.; Camper, L.; Buckley, J.; Pogue, E.; Banovac, K.

    2003-02-24T23:59:59.000Z

    On July 21, 1997, the U.S. Nuclear Regulatory Commission (NRC) published the final rule on Radiological Criteria for License Termination (the License Termination Rule or LTR) as Subpart E to 10 CFR Part 20. NRC regulations require that materials licensees submit Decommissioning Plans to support the decommissioning of its facility if it is required by license condition, or if the procedures and activities necessary to carry out the decommissioning have not been approved by NRC and these procedures could increase the potential health and safety impacts to the workers or the public. NRC regulations also require that reactor licensees submit Post-shutdown Decommissioning Activities Reports and License Termination Plans to support the decommissioning of nuclear power facilities. This paper provides an update on the status of the NRC's decommissioning program that was presented during WM'02. It discusses the staff's current efforts to streamline the decommissioning process, current issues being faced in the decommissioning program, such as partial site release and restricted release of sites, as well as the status of the decommissioning of complex sites and those listed in the Site Decommissioning Management Plan. The paper discusses the status of permanently shut-down commercial power reactors and the transfer of complex decommissioning sites and sites listed on the SDMP to Agreement States. Finally the paper provides an update of the status of various tools and guidance the NRC is developing to assist licensees during decommissioning, including an effort to consolidate and risk-inform decommissioning guidance.

  3. Twenty-first DOE/NRC nuclear air-cleaning conference

    SciTech Connect (OSTI)

    Bellamy, R.R. [Nuclear Regulatory Commission, Washington, DC (United States); Moeller, D.W.; First, M.W. [Harvard Univ., Cambridge, MA (United States)

    1991-01-01T23:59:59.000Z

    The Twenty-First Department of Energy/Nuclear Regulatory Commission Nuclear Air-Cleaning Conference was held August 12-16, 1990, in San Diego, California. A total of 232 air-cleaning specialists attended the conference. The United States and 14 foreign countries were represented, and the specialists were affiliated with government agencies, educational institutions, and the nuclear industry. Several major topics were discussed during the conference, including development and use of industry codes and standards; chemical processing off-gas cleaning; particulate filter developments, including filter testing and filter response to physical stress; development of adsorbents, including laboratory testing and in-place testing; incineration and vitrification; containment venting; reactor operations, including design and modeling; and measurement systems capable of verifying safe operation. The conference continued to provide a forum for direct and efficient interchange of technical and philosophical information among the participants. The high level of foreign participation and interest continues, as evidenced by over one half of the papers being sponsored by foreign interests, and one quarter of the attendees being from outside the United States. Further evidence of international interest was seen in a plenary session devoted to nuclear air-cleaning programs in nine different countries. A common concern throughout many of the sessions was the development of meaningful standards, their implementation for existing air-cleaning system, and the use of these standards by regulatory agencies. 13 refs., 2 tabs.

  4. STATUS OF THE NRC'S DECOMMISSIONING PROGRAM

    SciTech Connect (OSTI)

    Orlando, D. A.; Camper, L. W.; Buckley, J.

    2002-02-25T23:59:59.000Z

    On July 21, 1997, the U.S. Nuclear Regulatory Commission published the final rule on Radiological Criteria for License Termination (the License Termination Rule) as Subpart E to 10 CFR Part 20. NRC regulations require that materials licensees submit Decommissioning Plans to support the decommissioning of its facility if it is required by license condition, or if the procedures and activities necessary to carry out the decommissioning have not been approved by NRC and these procedures could increase the potential health and safety impacts to the workers or the public. NRC regulations also require that reactor licensees submit Post-shutdown Decommissioning Activities Reports and License Termination Plans to support the decommissioning of nuclear power facilities. This paper provides an update on the status of the NRC's decommissioning program. It discusses the status of permanently shut-down commercial power reactors, complex decommissioning sites, and sites listed in the Site Decommissioning Management Plan. The paper provides the status of various tools and guidance the NRC is developing to assist licensees during decommissioning, including a Standard Review Plan for evaluating plans and information submitted by licensees to support the decommissioning of nuclear facilities and the D and D Screen software for determining the potential doses from residual radioactivity. Finally, it discusses the status of the staff's current efforts to streamline the decommissioning process.

  5. NRC safety research in support of regulation - FY 1994. Volume 9

    SciTech Connect (OSTI)

    NONE

    1995-06-01T23:59:59.000Z

    This report, the tenth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1994. The goal of the Office of Nuclear Regulatory Research (RES) is to ensure the availability of sound technical bases for timely rulemaking and related decisions in support of NRC regulatory/licensing/inspection activities. RES also has responsibilities related to the resolution of generic safety issues and to the review of licensee submittals regarding individual plant examinations. It is the responsibility of RES to conduct the NRC`s rulemaking process, including the issuance of regulatory guides and rules that govern NRC licensed activities.

  6. Special committee review of the Nuclear Regulatory Commission's severe accident risks report (NUREG--1150)

    SciTech Connect (OSTI)

    Kouts, H.J.C. (Defense Nuclear Facility Safety Board (USA)); Apostolakis, G.; Kastenberg, W.E. (California Univ., Los Angeles, CA (USA)); Birkhofer, E.H.A. (Gesellschaft fuer Reaktorsicherheit mbH (GRS), Koeln (Germany, F.R.)); Hoegberg, L.G. (Swedish Nuclear Power Inspectorate, Stockholm (Sweden)); LeSage, L.G. (Argonne National Lab., IL (USA)); Rasmussen, N.C. (Massachusetts Inst. of Tech., Camb

    1990-08-01T23:59:59.000Z

    In April 1989, the Nuclear Regulatory Commission's (NRC) Office of Nuclear Regulatory Research (RES) published a draft report Severe Accident Risks: An Assessment for Five US Nuclear Power Plants,'' NUREG-1150. This report updated, extended and improved upon the information presented in the 1974 Reactor Safety Study,'' WASH-1400. Because the information in NUREG-1150 will play a significant role in implementing the NRC's Severe Accident Policy, its quality and credibility are of critical importance. Accordingly, the Commission requested that the RES conduct a peer review of NUREG-1150 to ensure that the methods, safety insights and conclusions presented are appropriate and adequately reflect the current state of knowledge with respect to reactor safety. To this end, RES formed a special committee in June of 1989 under the provisions of the Federal Advisory Committee Act. The Committee, composed of a group of recognized national and international experts in nuclear reactor safety, was charged with preparing a report reflecting their review of NUREG-1150 with respect to the adequacy of the methods, data, analysis and conclusions it set forth. The report which precedes reflects the results of this peer review.

  7. The U.S. Nuclear Regulatory Commission Thermal-Hydraulic Research Program: Maintaining expertise in a changing environment

    SciTech Connect (OSTI)

    Sheron, B.W.; Shotkin, L.M.; Baratta, A.J.

    1993-04-01T23:59:59.000Z

    Throughout the 1970s and early 1980s, the U.S. Nuclear Regulatory Commission`s (NRC`s) thermal-hydraulic research program enjoyed ample funding, sponsored extensive experimental and analytical development programs, and attracted worldwide expertise. With the completion of the major experimental programs and with the promulgation of the revised emergency core-cooling system rule, both the funding and prominence of thermal-hydraulic research at the NRC have declined in recent years. This has led justifiably to the concern by some that the program may no longer have the minimal elements needed to maintain both expertise and world-class status. The purpose of this article is to describe the NRC`s current thermal-hydraulic research program and to show how this program ensures maintenance of a viable, robust research effort and retention of needed expertise and international leadership.

  8. Recommendations for NEAMS Engagement with the NRC: Preliminary Report

    SciTech Connect (OSTI)

    Bernholdt, David E [ORNL

    2012-06-01T23:59:59.000Z

    The vision of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is to bring a new generation of analytic tools to the nuclear engineering community in order to facilitate students, faculty, industry and laboratory researchers in investigating advanced reactor and fuel cycle designs. Although primarily targeting at advance nuclear technologies, it is anticipated that these new capabilities will also become interesting and useful to the nuclear regulator Consequently, the NEAMS program needs to engage with the Nuclear Regulatory Commission as the software is being developed to ensure that they are familiar with and ready to respond to this novel approach when the need arises. Through discussions between key NEAMS and NRC staff members, we tentatively recommend annual briefings to the Division of Systems Analysis in the NRC's Office of Nuclear Regulatory Research. However the NEAC subcommittee review of the NEAMS program may yield recommendations that would need to be considered before finalizing this plan.

  9. Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report

    SciTech Connect (OSTI)

    NONE

    2000-08-01T23:59:59.000Z

    OAK B188 Risk-informed assessment of regulatory and design requirements for future nuclear power plants. Annual report. The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-formed approach for the design and regulation of nuclear power plants. This approach will include the development and/or confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRS) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go further by focusing on the design of new plants.

  10. Regulatory and technical reports (abstract index journal). Second quarter 1996, April--June

    SciTech Connect (OSTI)

    Sheehan, M.A.

    1996-08-01T23:59:59.000Z

    This report contains bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (US NRC) staff and it`s contractors.

  11. Regulatory and technical reports (abstract index journal): Annual compilation for 1996, Volume 21, No. 4

    SciTech Connect (OSTI)

    Sheehan, M.A.

    1997-04-01T23:59:59.000Z

    This compilation is the annual cumulation of bibliographic data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors.

  12. Development of the NRC`s Human Performance Investigation Process (HPIP). Volume 3, Development documentation

    SciTech Connect (OSTI)

    Paradies, M.; Unger, L. [System Improvements, Inc., Knoxville, TN (United States); Haas, P.; Terranova, M. [Concord Associates, Inc., Knoxville, TN (United States)

    1993-10-01T23:59:59.000Z

    The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause events at nuclear power plants. This document, Volume III, is a detailed documentation of the development effort and the pilot training program.

  13. Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2010, Prepared for the Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 2012

    SciTech Connect (OSTI)

    D. E. Lewis D. A. Hagemeyer Y. U. McCormick

    2012-07-07T23:59:59.000Z

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission’s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2010 annual reports submitted by five of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Because there are no geologic repositories for high-level waste currently licensed and no NRC-licensed low-level waste disposal facilities currently in operation, only five categories will be considered in this report. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Annual reports for 2010 were received from a total of 190 NRC licensees. The summation of reports submitted by the 190 licensees indicated that 192,424 individuals were monitored, 81,961 of whom received a measurable dose. When adjusted for transient workers who worked at more than one licensee during the year, there were actually 142,471 monitored individuals and 62,782 who received a measurable dose. The collective dose incurred by these individuals was 10,617 person-rem, which represents a 12% decrease from the 2009 value. This decrease was primarily due to the decrease in collective dose at commercial nuclear power reactors, as well as a decrease in the collective dose for most of the other categories of NRC licensees. The number of individuals receiving a measurable dose also decreased, resulting in an average measurable dose of 0.13 rem for 2010. The average measurable dose is defined as the total effective dose equivalent (TEDE) divided by the number of individuals receiving a measurable dose. In calendar year 2010, the average annual collective dose per reactor for light water reactor (LWR) licensees was 83 person-rem. This represents a 14% decrease from the value reported for 2009 (96 person-rem). The decrease in collective dose for commercial nuclear power reactors was due to an 11% decrease in total outage hours in 2010. During outages, activities involving increased radiation exposure such as refueling and maintenance are performed while the reactor is not in operation. The average annual collective dose per reactor for boiling water reactors (BWRs) was 137 personrem for 35 BWRs, and 55 person-rem for 69 pressurized water reactors (PWRs). Analyses of transient individual data indicate that 29,333 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient individuals by multiple licensees. The adjustment to account for transient individuals has been specifically noted in footnotes in the figures and tables for commercial nuclear power reactors. In 2010, the average measurable dose per individual for all licensees calculated from reported data was 0.13 rem. Although the average measurable dose per individual from data submitted by licensees was 0.13 rem, a corrected dose distribution resulted in an average measurable dose per individual of 0.17 rem.

  14. Reassessment of the NRC`s program for protecting allegers against retaliation

    SciTech Connect (OSTI)

    Not Available

    1994-01-01T23:59:59.000Z

    On July 6, 1993, the Nuclear Regulatory Commission`s (NRC`s) Executive Director for Operations established a review team to reassess the NRC`s program for protecting allegers against retaliation. The team evaluated the current system, and solicited comments from various NRC offices, other Federal agencies, licensees, former allegers, and the public. This report is subject to agency review. The report summarizes current processes and gives an overview of current problems. It discusses: (1) ways in which licensees can promote a quality-conscious work environment, in which all employees feel free to raise concerns without fear of retaliation; (2) ways to improve the NRC`s overall handling of allegations; (3) the NRC`s involvement in the Department of Labor process; (4) related NRC enforcement practices; and (5) methods other than investigation and enforcement that may be useful in treating allegations of potential or actual discrimination. Recommendations are given in each area.

  15. Report to the US Nuclear Regulatory Commission on Analysis and Evaluation of Operational Data, 1986

    SciTech Connect (OSTI)

    none,

    1987-05-01T23:59:59.000Z

    This annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during calendar year 1986. Comments and observations are provided on operating experience at nuclear power plants and other NRC licensees, including results from selected AEOD studies; summaries of abnormal occurrences involving US nuclear plants; reviews of licensee event reports and their quality, reactor scram experience from 1984 to 1986, engineered safety features actuations, and the trends and patterns analysis program; and assessments of nonreactor and medical misadministration events. In addition, the report provides the year-end status of all recommendations included in AEOD studies, and listings of all AEOD reports issued from 1980 through 1986.

  16. Nuclear Energy Research Initiative. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants. Annual Report

    SciTech Connect (OSTI)

    Ritterbusch, S.E.

    2000-08-01T23:59:59.000Z

    The overall goal of this research project is to support innovation in new nuclear power plant designs. This project is examining the implications, for future reactors and future safety regulation, of utilizing a new risk-informed regulatory system as a replacement for the current system. This innovation will be made possible through development of a scientific, highly risk-informed approach for the design and regulation of nuclear power plants. This approach will include the development and.lor confirmation of corresponding regulatory requirements and industry standards. The major impediment to long term competitiveness of new nuclear plants in the U.S. is the capital cost component--which may need to be reduced on the order of 35% to 40% for Advanced Light Water Reactors (ALWRs) such as System 80+ and Advanced Boiling Water Reactor (ABWR). The required cost reduction for an ALWR such as AP600 or AP1000 would be expected to be less. Such reductions in capital cost will require a fundamental reevaluation of the industry standards and regulatory bases under which nuclear plants are designed and licensed. Fortunately, there is now an increasing awareness that many of the existing regulatory requirements and industry standards are not significantly contributing to safety and reliability and, therefore, are unnecessarily adding to nuclear plant costs. Not only does this degrade the economic competitiveness of nuclear energy, it results in unnecessary costs to the American electricity consumer. While addressing these concerns, this research project will be coordinated with current efforts of industry and NRC to develop risk-informed, performance-based regulations that affect the operation of the existing nuclear plants; however, this project will go farther by focusing on the design of new plants.

  17. Congress, NRC mull utility access to FBI criminal files

    SciTech Connect (OSTI)

    Ultroska, D.

    1984-08-01T23:59:59.000Z

    Experiences at Alabama Power Company and other nuclear utilities have promped a request for institutionalizing security checks of personnel in order to eliminated convicted criminals and drug users. The Nuclear Regulatory Commission (NRC), which could provide FBI criminal history information by submitting fingerprints, does not do so, and would require new legislation to take on that duty. Believing that current malevolent employees can be managed with existing procedures, NRC allows criminal background checks only on prospective employees in order to avoid a negative social impact on personnel. Legislation to transfer criminal histories to nuclear facilities is now pending, and NRC is leaning toward a request for full disclosure, partly because of terrorist threats and partly to save manpower time and costs in reviewing case histories.

  18. Microsoft PowerPoint - 9_NRC Roundtable Cover_Wensday_NMMSS_2014...

    National Nuclear Security Administration (NNSA)

    and Industry Roundtable Discussion Paul Adam, Wolf Creek Nuclear Operating Corp Suzanne Ani, NRC Brian Horn, NRC...

  19. Nuclear Regulatory Commission issuances: Opinions and decisions of the Nuclear Regulatory Commission with selected orders, July 1--December 31, 1996. Volume 44, Pages 1--432

    SciTech Connect (OSTI)

    NONE

    1997-10-01T23:59:59.000Z

    The hardbound edition of the Nuclear Regulatory Commission Issuances is a final compilation of the monthly issuances. It includes all of the legal precedents for the agency within a six-month period. Any opinions, decisions, denials, memoranda and orders of the Commission inadvertently omitted from the monthly softbounds and any corrections submitted by the NRC legal staff to the printed softbound issuances are contained in the hardbound edition. Cross references in the text and indexes are to the NRCI page numbers which are the same as the page numbers in this publication. Issuances are referred to as follows: Commission--CLI, Atomic Safety and Licensing Boards--LBP, Administrative Law Judges--ALJ, Directors` Decisions--DD, and Decisions on Petitions for Rulemaking--DPRM.

  20. NRC Recommendations

    Broader source: Energy.gov [DOE]

    Presentation on NRC recommendations to the DOE Systems Analysis Workshop held in Washington, D.C. July 28-29, 2004 to discuss and define role of systems analysis in DOE Hydrogen Program.

  1. Regulatory and technical reports (abstract index journal): Annual compilation for 1994. Volume 19, Number 4

    SciTech Connect (OSTI)

    NONE

    1995-03-01T23:59:59.000Z

    This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the US Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC`s intention to publish this compilation quarterly and to cumulate it annually. The main citations and abstracts in this compilation are listed in NUREG number order. These precede the following indexes: secondary report number index, personal author index, subject index, NRC originating organization index (staff reports), NRC originating organization index (international agreements), NRC contract sponsor index (contractor reports), contractor index, international organization index, and licensed facility index. A detailed explanation of the entries precedes each index.

  2. Development of the NRC`s Human Performance Investigation Process (HPIP). Volume 2, Investigators`s Manual

    SciTech Connect (OSTI)

    Paradies, M.; Unger, L. [System Improvements, Inc., Knoxville, TN (United States); Haas, P.; Terranova, M. [Concord Associates, Inc., Knoxville, TN (United States)

    1993-10-01T23:59:59.000Z

    The three volumes of this report detail a standard investigation process for use by US Nuclear Regulatory Commission (NRC) personnel when investigating human performance related events at nuclear power plants. The process, called the Human Performance Investigation Process (HPIP), was developed to meet the special needs of NRC personnel, especially NRC resident and regional inspectors. HPIP is a systematic investigation process combining current procedures and field practices, expert experience, NRC human performance research, and applicable investigation techniques. The process is easy to learn and helps NRC personnel perform better field investigations of the root causes of human performance problems. The human performance data gathered through such investigations provides a better understanding of the human performance issues that cause event at nuclear power plants. This document, Volume II, is a field manual for use by investigators when performing event investigations. Volume II includes the HPIP Procedure, the HPIP Modules, and Appendices that provide extensive documentation of each investigation technique.

  3. Nuclear Regulatory Commission issuances, January 1997. Volume 45, Number 1

    SciTech Connect (OSTI)

    NONE

    1997-01-01T23:59:59.000Z

    This book contains issuances of the Atomic Safety and Licensing Board, Nuclear Regulatory Commission and Director`s Decision for January 1997. The issuances concern Sequoyah Fuels Corporation and General Atomics Gore, Oklahoma Site decontamination and decommissioning funding; Louisiana Energy Services, Claiborne Enrichment Center denies appeal to review emergency planning; General Public Utilities Nuclear Corporation, Oyster Creek Nuclear Generating station, challenges to technical specifications concerning spent fuel pool; and Consumers Power Company, Palisades Nuclear Plant dry cask storage of spent nuclear fuel.

  4. Proceedings of the 21st DOE/NRC Nuclear Air Cleaning Conference; Sessions 1--8

    SciTech Connect (OSTI)

    First, M.W. [ed.] [Harvard Univ., Boston, MA (United States). Harvard Air Cleaning Lab.

    1991-02-01T23:59:59.000Z

    Separate abstracts have been prepared for the papers presented at the meeting on nuclear facility air cleaning technology in the following specific areas of interest: air cleaning technologies for the management and disposal of radioactive wastes; Canadian waste management program; radiological health effects models for nuclear power plant accident consequence analysis; filter testing; US standard codes on nuclear air and gas treatment; European community nuclear codes and standards; chemical processing off-gas cleaning; incineration and vitrification; adsorbents; nuclear codes and standards; mathematical modeling techniques; filter technology; safety; containment system venting; and nuclear air cleaning programs around the world. (MB)

  5. Uranium recovery research sponsored by the Nuclear Regulatory Commission at Pacific Northwest Laboratory. Annual progress report, May 1982-May 1983

    SciTech Connect (OSTI)

    Foley, M.G.; Opitz, B.E.; Deutsch, W.J.; Peterson, S.R.; Gee, G.W.; Serne, R.J.; Hartley, J.N.; Thomas, V.W.; Kalkwarf, D.R.; Walters, W.H.

    1983-06-01T23:59:59.000Z

    Pacific Northwest Laboratory (PNL) is currently conducting research for the US Nuclear Regulatory Commission (NRC) on uranium recovery process wastes for both active and inactive operations. NRC-sponsored uranium recovery research at PNL is focused on NRC regulatory responsibilities for uranium-recovery operations: license active milling and in situ extraction operations; concur on the acceptability of DOE remedial-action plans for inactive sites; and license DOE to maintain inactive sites following remedial actions. PNL's program consists of four coordinated projects comprised of a program management task and nine research tasks that address the critical technical and safety issues for uranium recovery. Specifically, the projects endeavor to find and evaluate methods to: prevent erosion of tailings piles and prevent radon release from tailings piles; evaluate the effectiveness of interim stabilization techniques to prevent wind erosion and transport of dry tailings from active piles; estimate the dewatering and consolidation behavior of slurried tailings to promote early cover placement; design a cover-protection system to prevent erosion of the cover by expected environmental stresses; reduce seepage into ground water and prevent ground-water degradation; control solution movement and reaction with ground water in in-situ extraction operations; evaluate natural and induced restoration of ground water in in-situ extraction operations; and monitor releases to the environment from uranium recovery facilities.

  6. Regulatory practices in India for establishing nuclear power stations

    SciTech Connect (OSTI)

    De, A.K. [Atomic Energy Regulatory Board, Calcutta (India); Singh, S.P. [Atomic Energy Regulatory Board, Bombay (India)

    1991-07-01T23:59:59.000Z

    The Atomic Energy Regulatory Board (AERB) of India was established as an independent regulatory authority charged with regulating radiation protection and nuclear safety. This article reviews the current state of India`s nuclear power reactor program and discusses the makeup of functions of the AERB, including the preparation of issuance of safety codes, guides, and other standards, with special recent emphasis on pressurized-heavy-water reactors (PHWRs). The AERB`s relationship to nuclear plant owners is discussed, as are the inspection and control functions the AERB performs, both for the construction and operation of nuclear plants and the licensing of operating personnel. 8 refs., 2 figs.

  7. Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview

    SciTech Connect (OSTI)

    Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

    1982-02-01T23:59:59.000Z

    In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

  8. U.S. Nuclear Regulatory Commission accountability report, fiscal year 1995. Volume 1

    SciTech Connect (OSTI)

    NONE

    1996-05-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) is one of six Federal agencies participating in a pilot project to streamline financial management reporting. The goal of this pilot is to consolidate performance-related reporting into a single accountability report. The project, which is being carried out under the guidance of the Chief Financial Officers Council, was undertaken in accordance with the Government Management Reform Act (GMRA) of 1994. The GMRA permits the streamlining of financial management reports in consultation with the appropriate Congressional Committees through a liaison in the US Office of Management and Budget (OMB). The results of the pilot project will determine the method to be used for reporting financial management information for fiscal year (FY) 1996. This report consolidates the information previously reported in the following documents: (1) the NRC`s annual financial statement required by the Chief Financial Officers Act of 1990; (2) the Chairman`s annual report to the President and the Congress, required by the Federal Managers` Financial Integrity Act of 1982; (3) the Chairman`s semiannual report to the Congress on management decisions and final actions on Office of Inspector General audit recommendations, required by the Inspector General Act of 1978, as amended. This report also includes performance measures, as required by the Chief Financial Officers Act of 1990.

  9. NRC policy on future reactor designs

    SciTech Connect (OSTI)

    none,

    1985-07-01T23:59:59.000Z

    On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions.

  10. argentine nuclear regulatory: Topics by E-print Network

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

    argentine nuclear regulatory 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 UNITED STATES NUCLEAR...

  11. UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

    E-Print Network [OSTI]

    Laughlin, Robert B.

    In the Matter of CALVERT CLIFFS 3 NUCLEAR PROJECT, LLC, and UNISTAR NUCLEAR OPERATING SERVICES, LLC (Combined, 2009 MEMORANDUM AND ORDER (Ruling on Joint Petitioners' Standing and Contentions) I. Introduction This case arises from an application by UniStar Nuclear Operating Services, LLC and Calvert Cliffs 3 Nuclear

  12. Recommendations to the NRC on human engineering guidelines for nuclear power plant maintainability

    SciTech Connect (OSTI)

    Badalamente, R.V.; Fecht, B.A.; Blahnik, D.E.; Eklund, J.D.; Hartley, C.S.

    1986-03-01T23:59:59.000Z

    This document contains human engineering guidelines which can enhance the maintainability of nuclear power plants. The guidelines have been derived from general human engineering design principles, criteria, and data. The guidelines may be applied to existing plants as well as to plants under construction. They apply to nuclear power plant systems, equipment and facilities, as well as to maintenance tools and equipment. The guidelines are grouped into seven categories: accessibility and workspace, physical environment, loads and forces, maintenance facilities, maintenance tools and equipment, operating equipment design, and information needs. Each chapter of the document details specific maintainability problems encountered at nuclear power plants, the safety impact of these problems, and the specific maintainability design guidelines whose application can serve to avoid these problems in new or existing plants.

  13. Comparative Analysis Between US NRC Requirements and US DOE Orders - 13402

    SciTech Connect (OSTI)

    Chakraborti, Sayan [MRIGlobal, 425 Volker Blvd, Kansas City, MO 64110 (United States)] [MRIGlobal, 425 Volker Blvd, Kansas City, MO 64110 (United States); Stone, Lynn; Hyatt, Jeannette [Savannah River Nuclear Solutions (United States)] [Savannah River Nuclear Solutions (United States)

    2013-07-01T23:59:59.000Z

    Small modular reactor (SMR) is a nuclear reactor design approach that is expected to herald in a new era of clean energy in the U.S. These reactors are less than one-third the size of conventional large nuclear power reactors, and have factory-fabricated components that may be transported by rail or truck to a site selected to house a small nuclear reactor. To facilitate the licensing of these smaller nuclear reactor designs, the Nuclear Regulatory Commission (NRC) is in the process of developing a regulatory infrastructure to support licensing review of these unique reactor designs. As part of these activities, the NRC has been meeting with the Department of Energy (DOE) and with individual SMR designers to discuss potential policy, licensing, and key technical differences in SMR designs. It is anticipated by the NRC that such licensing interaction and guidance early in the design process will contribute towards minimizing complexity while adding stability and predictability in the licensing and subsequent regulation of new reactor designs such as SMRs. In conjunction with the current NRC initiative of developing the SMR licensing process, early communication and collaboration in the identification and resolution of any potential technical and licensing differences between NRC requirements and similar requirements applicable at DOE sites would help to expedite demonstration and implementation of SMR technology in the US. In order to foster such early communication, Savannah River Nuclear Solutions (SRNS) has begun taking the first steps in identifying and evaluating potential licensing gaps that may exist between NRC and DOE requirements in siting SMRs at DOE sites. A comparison between the existing NRC regulations for Early Site Permits and the DOE Orders was undertaken to establish the degree of correlation between NRC requirements and compliance methods in place at DOE sites. The ability to use existing data and information to expedite the development of the Environmental Report is being evaluated at the Savannah River Site as a case study for application across the DOE Complex. This paper will present areas of direct correlation as well as those where the need for site specific data for either DOE operations or NRC compliance warrant additional interaction between the agencies. Areas where further refinement of the SMR technologies may drive collaborative development of revised regulations through such means as industry consensus standards will also be highlighted. Both NRC and DOE have requirements that mandate public involvement in their processes. The importance and value of early engagement with the public as well as collaborating regulatory agencies is of critical importance when deploying new technologies. (authors)

  14. Coordinating NRC License Closure/Termination and Army Corps of Engineers FUSRAP Cleanups

    SciTech Connect (OSTI)

    Walter, N. [MACTEC, 511 Congress Street, Portland, ME 04101 (United States); Greene, D. R. [LeBoeuf, Lamb, Greene and MacRae LLP, 225 Asylum Street, Hartford, CT 06103 (United States); Knauerhase, R. K. [Combustion Engineering, 2000 Day Hill Road, CEP 5580-2207, Windsor, CT 06095 (United States)

    2006-07-01T23:59:59.000Z

    Overlapping regulatory cleanup programs present a significant challenge for business entities seeking to close and redevelop properties in an environmentally-appropriate but cost-effective manner. In the nuclear decontamination context, this challenge has been recognized in Memoranda of Understanding ('MOUs') between regulators with overlapping responsibilities seeking to minimize duplicative efforts/costs while fulfilling their respective regulatory obligations. For instance, an MOU between the Army Corps of Engineers (the 'Corps') and the Nuclear Regulatory Commission ('NRC') for coordinating Corps' cleanups under the Formerly Utilized Sites Remedial Action Program ('FUSRAP') and NRC D and D to close and terminate an NRC license was reached in July 2001. Similarly, U.S. Environmental Protection Agency ('EPA') and NRC entered into an MOU in October 2002 addressing the interaction between NRC decontamination and decommissioning ('D and D') oversight and EPA's authority under the Comprehensive Environmental Response, Compensation and Liability Act ('CERCLA') at NRC-licensed sites. Yet, despite these MOU agreements, the simultaneous application of different regulatory programs, differing perspectives on their respective objectives and limited experience in addressing such circumstances often can lead to issues that demand creative solutions. This paper examines the interplay of these regulatory programs, the MOU of the agencies seeking to address their responsibilities under them and the coordination of the cleanups and license closure/termination process under the programs. It also offers technical and practical suggestions and insight to cost-effectively manage such efforts based on experiences with these programs and the regulators and stakeholders involved (at the federal, state and local levels). (authors)

  15. NRC support for the Kalinin (VVER) probabilistic risk assessment

    SciTech Connect (OSTI)

    Bley, D. [Buttonwood Consulting, Inc. (United States); Diamond, D.J.; Chu, T.L.; Azarm, A.; Pratt, W.T. [Brookhaven National Lab., Upton, NY (United States); Johnson, D. [PLG, Inc. (United States); Szukiewicz, A.; Drouin, M.; El-Bassioni, A.; Su, T.M. [Nuclear Regulatory Commission (United States)

    1998-12-31T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA.

  16. NRC SUPPORT FOR THE KALININ (VVER) PROBABILISTIC RISK ASSESSMENT

    SciTech Connect (OSTI)

    BLEY,D.; DIAMOND,D.J.; CHU,T.L.; AZARM,A.; PRATT,W.T.; JOHNSON,D.; SZUKIEWICZ,A.; DROUIN,M.; EL-BASSIONI,A.; SU,T.M.

    1998-10-26T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA.

  17. NUCLEAR REGULATORY,.COMMISSION REGION I

    Office of Legacy Management (LM)

    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 group currentBradleyTableSelling7 August 2008REGULATORY,.COMMISSION REGION I

  18. Comparison and Analysis of Regulatory and Derived Requirements for Certain DOE Spent Nuclear Fuel Shipments; Lessons Learned for Future Spent Fuel Transportation Campaigns

    SciTech Connect (OSTI)

    Kramer, George L., Ph.D.; Fawcett, Rick L.; Rieke, Philip C.

    2003-02-27T23:59:59.000Z

    Radioactive materials transportation is stringently regulated by the Department of Transportation and the Nuclear Regulatory Commission to protect the public and the environment. As a Federal agency, however, the U.S. Department of Energy (DOE) must seek State, Tribal and local input on safety issues for certain transportation activities. This interaction has invariably resulted in the imposition of extra-regulatory requirements, greatly increasing transportation costs and delaying schedules while not significantly enhancing the level of safety. This paper discusses the results an analysis of the regulatory and negotiated requirements established for a July 1998 shipment of spent nuclear fuel from foreign countries through the west coast to the Idaho National Engineering and Environmental Laboratory (INEEL). Staff from the INEEL Nuclear Materials Engineering and Disposition Department undertook the analysis in partnership with HMTC, to discover if there were instances where requirements derived from stakeholder interactions duplicate, contradict, or otherwise overlap with regulatory requirements. The study exhaustively lists and classifies applicable Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) regulations. These are then compared with a similarly classified list of requirements from the Environmental Impact Statements (EIS) and those developed during stakeholder negotiations. Comparison and analysis reveals numerous attempts to reduce transportation risk by imposing more stringent safety measures than those required by DOT and NRC. These usually took the form of additional inspection, notification and planning requirements. There are also many instances of overlap with, and duplication of regulations. Participants will gain a greater appreciation for the need to understand the risk-oriented basis of the radioactive materials regulations and their effectiveness in ensuring safety when negotiating extra-regulatory requirements.

  19. Report to the US Nuclear Regulatory Commission on analysis and evaluation of operational data - 1987: Power reactors

    SciTech Connect (OSTI)

    none,

    1988-10-01T23:59:59.000Z

    This annual report of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) is devoted to the activities performed during 1987. The report is published in two volumes. NUREG-1272, Vol. 2, No. 1, covers Power Reactors and presents an overview of the operating experience of the nuclear power industry, with comments regarding the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year, and summarizes information from Licensee Event Reports, the NRC's Operations Center, and Diagnostic Evaluations. NUREG-1272, Vol. 2, No. 2, covers Nonreactors and presents a review of the nonreactors events and misadministration reports that were reported in 1987 and a brief synopsis of AEOD studies published in 1987. Each volume contains a list of the AEOD Reports issued for 1980-1987.

  20. Government perspective. [FY 1994 budgets for DOE and NRC

    SciTech Connect (OSTI)

    Not Available

    1993-05-01T23:59:59.000Z

    During April, several Congressional committees held hearings on the FY 1994 budget submitted by the Department of Energy (DOE). DOE has requested a total nuclear energy budget of $165 million, reduced from $299.3 million in FY 1993. Energy Secretary Hazel O'Leary and acting Assistant Secretary for Nuclear Energy, Mr. E.C. Brolin, testified that DOE intends to continue to fund research for light water reactors (LWR) at a level of $57.8 million during FY 1994, a slight decline from the $58.7 million requested for FY 1993. The LWR program funding will allow work to continue on the standardization and design certification programs. However, DOE's proposed $45 million budget cut, to only $15 million, for funding of advanced nuclear research and development programs has been sharply criticized by nuclear industry supporters in Congress. The Nuclear Regulatory Commission's (NRC) Chairman, Ivan Selin, testified before the House Energy and Water Appropriations Subcommittee regarding NRC's budget request. NRC is requesting $547.7 million for FY 1994, an increase of 1.5 percent over the FY 1993 budget.

  1. 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

  2. HFE Process Guidance and Standards for potential application to updating NRC guidance

    SciTech Connect (OSTI)

    Jacques Hugo; J. J. Persensky

    2012-07-01T23:59:59.000Z

    The U.S. Nuclear Regulatory Commission (NRC) reviews and evaluates 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 these guidance documents to ensure that they remain state-of-the-art design evaluation tools. Thus, the NRC has initiated a project with BNL to update the NRC guidance to remain current with recent research on human performance, advances in HFE methods and tools, and new technology. INL supported Brookhaven National Lab (BNL) to update the detailed HFE review criteria contained in NUREG-0711 and NUREG-0700 based on (1) feedback obtained from end users, (2) the results of NRC research and development efforts supporting the NRC staff’s HFE safety reviews, and (3) other material the project staff identify as applicable to the update effort. INL submitted comments on development plans and sections of NUREGs 0800, 0711, and 0700. The contractor prepared the report attached here as the deliverable for this work.

  3. Regulatory Oversight Program, July 1, 1993--March 3, 1997. Volume 2: Appendices

    SciTech Connect (OSTI)

    NONE

    1997-12-31T23:59:59.000Z

    On July, 1993, a Regulatory Oversight (RO) organization was established within the US DOE, Oak Ridge Operations to provide regulatory oversight of the DOE uranium enrichment facilities leased to the United States Enrichment Corporation (USEC). The purpose of the OR program was to ensure continued plant safety, safeguards and security while the plants were transitioned to regulatory oversight by the Nuclear Regulatory Commission (NRC). Volume 2 contains copies of the documents which established the relationship between NRC, DOE, USEC, and DOL (Dept of Labor) required to facilitate regulatory oversight transition.

  4. New security measures are proposed for N-plants: Insider Rule package is issued by NRC

    SciTech Connect (OSTI)

    Not Available

    1984-09-01T23:59:59.000Z

    New rules proposed by the Nuclear Regulatory Commission (NRC) will require background investigations and psychological assessments of new job candidates and continual monitoring of the behavior of all power plant workers with access to sensitive areas. Licensees will have to submit an ''access authorization'' program for approval describing how they will conduct these security activities. The employee checks will go back five years to examine credit, educational, and criminal histories. Implementation of the rules could involve the Edison Electric Institute as an intermediary to funnel criminal checks from the Justice Department and FBI. The NRC is also considering a clarification of areas designated as ''vital'' because current designations may be too strict.

  5. Selection of a discount rate for use in NRC regulatory analyses and application of discount rates to future averted health effects

    SciTech Connect (OSTI)

    Paananen, O.H.; Hendrickson, P.L.

    1993-01-01T23:59:59.000Z

    The principal objective of this report is to provide background information and recommendations on the use of discount rates in the regulatory analysis process. The report focuses on two issues selecting the appropriate discount rate or rates to use when conducting a regulatory analysis, and applying the selected discount rate to future health-related benefits estimated to result from alternative regulatory actions.

  6. IN NRC PUBLICATIONS NRC Reference Material

    E-Print Network [OSTI]

    Decommissioning Process

    1999-01-01T23:59:59.000Z

    Publicly released records include, to name a few, NUREG-series publications; Federal Register notices; applicant, licensee, and vendor documents and correspondence; NRC correspondence and internal memoranda; bulletins and information notices; inspection and investigative reports; licensee event reports; and Commission papers and their attachments. NRC publications in the NUREG series, NRC regulations, and Title 10, Energy, in the Code of Federal Regulations may also be purchased from one of these two sources.

  7. NRC drug-free workplace plan. Revision 1

    SciTech Connect (OSTI)

    NONE

    1997-11-01T23:59:59.000Z

    On September 15, 1986, President Reagan signed Executive Order 12564, establishing the goal of a Drug-Free Federal Workplace. The Order made it a condition of employment that all Federal employees refrain from using illegal drugs on or off duty. On July 11, 1987, Congress passed legislation affecting implementation of the Executive Order under Section 503 of the Supplemental Appropriations Act of 1987, Public Law 100-71 (the Act). The Nuclear Regulatory Commission first issued the NRC Drug Testing Plan to set forth objectives, policies, procedures, and implementation guidelines to achieve a drug-free Federal workplace, consistent with the Executive Order and Section 503 of the Act. Revision 1, titled, ``NRC Drug-Free Workplace Plan,`` supersedes the previous version and its supplements and incorporates changes to reflect current guidance from the Department of Justice, the Department of Health and Human Services, as well as other guidance.

  8. NRC Job Code V6060: Extended in-situ and real time monitoring. Task 4: Detection and monitoring of leaks at nuclear power plants external to structures

    SciTech Connect (OSTI)

    Sheen, S. H. (Nuclear Engineering Division)

    2012-08-01T23:59:59.000Z

    In support of Task 4 of the NRC study on compliance with 10 CFR part 20.1406, minimization of contamination, Argonne National Laboratory (ANL) conducted a one-year scoping study, in concert with a parallel study performed by NRC/NRR staff, on monitoring for leaks at nuclear power plants (NPPs) external to structures. The objective of this task-4 study is to identify and assess those sensors and monitoring techniques for early detection of abnormal radioactive releases from the engineered facility structures, systems and components (SSCs) to the surrounding underground environment in existing NPPs and planned new reactors. As such, methods of interest include: (1) detection of anomalous water content of soils surrounding SSCs, (2) radionuclides contained in the leaking water, and (3) secondary signals such as temperature. ANL work scope includes mainly to (1) identify, in concert with the nuclear industry, the sensors and techniques that have most promise to detect radionuclides and/or associated chemical releases from SSCs of existing NPPs and (2) review and provide comments on the results of the NRC/NRR staff scoping study to identify candidate technologies. This report constitutes the ANL deliverable of the task-4 study. It covers a survey of sensor technologies and leak detection methods currently applied to leak monitoring at NPPs. The survey also provides a technology evaluation that identifies their strength and deficiency based on their detection speed, sensitivity, range and reliability. Emerging advanced technologies that are potentially capable of locating releases, identifying the radionuclides, and estimating their concentrations and distributions are also included in the report along with suggestions of required further research and development.

  9. Opinions and decisions of the Nuclear Regulatory Commission with selected orders, July 1, 1995--December 31, 1995. Volume 42, Pages 1-258

    SciTech Connect (OSTI)

    NONE

    1996-11-01T23:59:59.000Z

    This is the 42nd volume of issuances of the U.S. Nuclear Regulatory Commission (NRC) and its Atomic Safety and Licensing Boards, Administrative Law Judges, and Office Directors. This book is a reprinting, containing corrections of numerous printing errors in a previously distributed book. It covers the period from July 1, 1995 to December 31, 1995. Atomic Safety and Licensing Boards conduct adjudicatory hearings on applications to construct and operate nuclear power plants and related facilities, and issue initial decisions which, subject to internal review and appellate procedures, become the final Commission action with respect to those applications. The hardbound edition of the Nuclear Regulatory Commission Issuances is a final compilation of the monthly issuances. It includes all of the legal precedents for the agency within a 6-month period. Any opinions, decisions, denials, memoranda and orders of the Commission inadvertently omitted from the monthly editions and any corrections submitted by the NRC legal staff to the printed softbound issuances are contained in the hardbound edition.

  10. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    SciTech Connect (OSTI)

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09T23:59:59.000Z

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  11. 4th Annual NRC Small Business Seminar and Matchmaking Event

    Broader source: Energy.gov [DOE]

    The U.S. NRC regulates commercial and institutional uses of nuclear energy to protect the public health and safety, promote the common defense and security, and protect the environment. Attendees will hear directly from NRC leadership on the vision and priorities of the NRC at its Headquarters in Rockville, Maryland, including current acquisition initiatives and how small businesses can support their mission.

  12. Approaches used for Clearance of Lands from Nuclear Facilities among Several Countries: Evaluation for Regulatory Input

    Broader source: Energy.gov [DOE]

    The study entitled, “Approaches used for Clearance of Lands from Nuclear Facilities among Several Countries: Evaluation for Regulatory Input,” focuses on the issue of showing compliance with given...

  13. U.S. DOE Approach to Address U.S. NRC Key Technical Issues for a License Application

    SciTech Connect (OSTI)

    Ziegler, J. D.; Gunter, T. C.; Gamble, R. P.; Bradbury, R. B.

    2003-02-25T23:59:59.000Z

    Interactions between the U.S. Department of Energy (DOE) staff and the U.S. Nuclear Regulatory Commission (NRC) staff prior to submittal of a license application (LA) for NRC review are focused on resolution of issues relevant to licensing a geologic repository at the Yucca Mountain site. These interactions take place in meetings that are open to the public, the State of Nevada, affected units of local government, and other interested parties. Consistent with a 1992 agreement between the DOE and NRC, resolution of an issue at the staff level can be achieved during the pre-licensing period when the NRC staff has no further questions or comments regarding how the DOE is addressing that issue. In no case does such resolution at the NRC staff level preclude an issue being raised during the licensing proceedings by the NRC or another party to the proceedings. Beginning in 1996, interactions between the DOE and NRC began to focus significant attention on the nine topical areas, called Key Technical Issues (KTIs), that the NRC staff considers to be important in evaluating the post-closure performance of a Yucca Mountain repository. DOENRC meetings to discuss each KTI and achieve technical agreement on the information needed to resolve the issues were held between August 2000 and September 2001. As a result of these meetings, 293 agreements were reached regarding information to be developed by the DOE to supplement the basis for NRC review of the initial LA.* As of April 23, 2003, 77 of these agreements are considered by the NRC to be complete based on information provided by the DOE.

  14. The future of nuclear power in the United States : economic and regulatory challenges

    E-Print Network [OSTI]

    Joskow, Paul L.

    2006-01-01T23:59:59.000Z

    This paper examines the economic and regulatory challenges that must be faced by potential investors in new nuclear power plants in the United States. The historical development of the existing fleet of over 100 nuclear ...

  15. Summary of Information Presented at an NRC-Sponsored Low-Power Shutdown Public Workshop, April 27, 1999, Rockville, Maryland

    SciTech Connect (OSTI)

    Wheeler, Timothy A.; Whitehead, Donnie W.; Lois, Erasmia

    1999-07-01T23:59:59.000Z

    This report summarizes a public workshop that was held on April 27, 1999, in Rockville, Maryland. The workshop was conducted as part of the US Nuclear Regulatory Commission's (NRC) efforts to further develop its understanding of the risks associated with low power and shutdown operations at US nuclear power plants. A sufficient understanding of such risks is required to support decision-making for risk-informed regulation, in particular Regulatory Guide 1.174, and the development of a consensus standard. During the workshop the NRC staff discussed and requested feedback from the public (including representatives of the nuclear industry, state governments, consultants, private industry, and the media) on the risk associated with low-power and shutdown operations.

  16. Nuclear Regulatory Commission issuances, April 1995. Volume 41, Number 4

    SciTech Connect (OSTI)

    NONE

    1995-04-01T23:59:59.000Z

    This book contains issuances of the Nuclear Regulatory Commission and of the Atomic Safety and Licensing Boards, and an issuance of the Director`s decision. The issuances concern a petition filed by Dr. James E Bauer seeking interlocutory Commission review of the Atomic Safety and Licensing Board`s order imposing several restrictions on Dr. Bauer; a denial of an Interveners` Petition for Review addressing the application of Babcock and Wilcox for a renewal of its Special Nuclear Materials License; granting a motion for a protective order, by Sequoyah Fuel Corporation and General Atomics, limiting the use of the protected information to those individuals participating in the litigation and for the purposes of the litigation only; granting a Petitioner`s petition for leave to intervene and request for a hearing concerning Georgia Institute of Technology (Georgia Tech Research Reactor) renewal of a facility license; and a denial of a petition filed by Mr. Ted Dougherty requesting a shutdown of the San Onofre Nuclear Generating Station based on concerns regarding the vulnerability of the plant to earthquakes and defensibility of the plant to a terrorist threat.

  17. A review of NRC staff uses of probabilistic risk assessment

    SciTech Connect (OSTI)

    Not Available

    1994-03-01T23:59:59.000Z

    The NRC staff uses probabilistic risk assessment (PRA) and risk management as important elements its licensing and regulatory processes. In October 1991, the NRC`s Executive Director for Operations established the PRA Working Group to address concerns identified by the Advisory Committee on Reactor Safeguards with respect to unevenness and inconsistency in the staff`s current uses of PRA. After surveying current staff uses of PRA and identifying needed improvements, the Working Group defined a set of basic principles for staff PRA use and identified three areas for improvements: guidance development, training enhancements, and PRA methods development. For each area of improvement, the Working Group took certain actions and recommended additional work. The Working Group recommended integrating its work with other recent PRA-related activities the staff completed and improving staff interactions with PRA users in the nuclear industry. The Working Group took two key actions by developing general guidance for two uses of PRA within the NRC (that is, screening or prioritizing reactor safety issues and analyzing such issues in detail) and developing guidance on basic terms and methods important to the staff`s uses of PRA.

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

    SciTech Connect (OSTI)

    Travis, R.J.; Davis, R.E.; Grove, E.J.; Azarm, M.A. [Brookhaven National Lab., Upton, NY (United States)

    1997-08-01T23:59:59.000Z

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

  19. Control and Accountability of Nuclear Materials

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

    1999-08-11T23:59:59.000Z

    DOE O 474.1 prescribes Department of Energy (DOE) requirements for nuclear material control and accountability (MC&A) for DOE-owned and -leased facilities and DOE-owned nuclear materials at other facilities which are exempt from licensing by the Nuclear Regulatory Commission (NRC). Cancels DOE 5633.3B

  20. Control and Accountability of Nuclear Materials

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

    1993-02-12T23:59:59.000Z

    The order prescribes DOE minimum requirements and procedures for control and accountability of nuclear materials at DOE-owned and -leased facilities and DOE-owned nuclear materials at other facilities which are exempt from licensing by the Nuclear Regulatory Commission {NRC). Cancels DOE O 5633.3. Canceled by DOE O 5633.3B.

  1. 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

  2. Codes and standards and other guidance cited in regulatory documents. Revision 1

    SciTech Connect (OSTI)

    Ankrum, A.; Nickolaus, J.; Vinther, R.; Maguire-Moffitt, N.; Hammer, J.; Sherfey, L.; Warner, R. [Pacific Northwest Lab., Richland, WA (United States)

    1994-08-01T23:59:59.000Z

    As part of the US Nuclear Regulatory Commission (NRC) Standard Review Plan Update and Development Program, Pacific Northwest Laboratory developed a listing of industry consensus codes and standards and other government and industry guidance referred to in regulatory documents. In addition to updating previous information, Revision 1 adds citations from the NRC Inspection Manual and the Improved Standard Technical Specifications. This listing identifies the version of the code or standard cited in the regulatory document, the regulatory document, and the current version of the code or standard. It also provides a summary characterization of the nature of the citation. This listing was developed from electronic searches of the Code of Federal Regulations and the NRC`s Bulletins, Information Notices, Circulars, Generic Letters, Policy Statements, Regulatory Guides, and the Standard Review Plan (NUREG-0800).

  3. Regulatory analysis technical evaluation handbook. Final report

    SciTech Connect (OSTI)

    NONE

    1997-01-01T23:59:59.000Z

    The purpose of this Handbook is to provide guidance to the regulatory analyst to promote preparation of quality regulatory analysis documents and to implement the policies of the Regulatory Analysis Guidelines of the US Nuclear Regulatory Commission (NUREG/BR-0058 Rev. 2). This Handbook expands upon policy concepts included in the NRC Guidelines and translates the six steps in preparing regulatory analyses into implementable methodologies for the analyst. It provides standardized methods of preparation and presentation of regulatory analyses, with the inclusion of input that will satisfy all backfit requirements and requirements of NRC`s Committee to Review Generic Requirements. Information on the objectives of the safety goal evaluation process and potential data sources for preparing a safety goal evaluation is also included. Consistent application of the methods provided here will result in more directly comparable analyses, thus aiding decision-makers in evaluating and comparing various regulatory actions. The handbook is being issued in loose-leaf format to facilitate revisions. NRC intends to periodically revise the handbook as new and improved guidance, data, and methods become available.

  4. US Nuclear Regulatory Commission organization charts and functional statements

    SciTech Connect (OSTI)

    NONE

    1996-07-01T23:59:59.000Z

    This document is the US NRC organizational structure and chart as of July 1, 1996. It contains the org charts for the Commission, ACRS, ASLAB, Commission staff offices, Executive Director for Operations, Office of the Inspector General, Program offices, and regional offices.

  5. Regulatory and technical reports (abstract index journal): Compilation for third quarter 1994, July--September. Volume 19, Number 3

    SciTech Connect (OSTI)

    none,

    1994-12-01T23:59:59.000Z

    This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issues by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC`s intention to publish this compilation quarterly and to cumulate it annually. The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/IA-XXXX. These precede the following indexes: Secondary Report Number Index, Personal Author Index, Subject Index, NRC Originating Organization Index (Staff Reports), NRC Originating Organization Index (International Agreements), NRC Contract Sponsor Index (Contractor Reports) Contractor Index, International Organization Index, Licensed Facility Index. A detailed explanation of the entries precedes each index.

  6. Regulatory and technical reports (abstract index journal): Compilation for third quarter 1996 July--September. Volume 21, Number 3

    SciTech Connect (OSTI)

    NONE

    1997-02-01T23:59:59.000Z

    This compilation consists of bibliographic data and abstracts for the formal regulatory and technical reports issued by the US Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC`s intention to publish this compilation quarterly and to cumulate it annually. The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREG/CP-XXXX, NUREG/CR-XXXX, and NUREG/IA-XXXX. These precede the following indexes: secondary report number index; personal author index; subject index; NRC originating organization index (staff reports); NRC originating organization index (international agreements); NRC contract sponsor index (contractor reports); contractor index; international organization index; and licensed facility index. A detailed explanation of the entries precedes each index.

  7. Peer Review of NRC Standardized Plant Analysis Risk Models

    SciTech Connect (OSTI)

    Anthony Koonce; James Knudsen; Robert Buell

    2011-03-01T23:59:59.000Z

    The Nuclear Regulatory Commission (NRC) Standardized Plant Analysis Risk (SPAR) Models underwent a Peer Review using ASME PRA standard (Addendum C) as endorsed by NRC in Regulatory Guide (RG) 1.200. The review was performed by a mix of industry probabilistic risk analysis (PRA) experts and NRC PRA experts. Representative SPAR models, one PWR and one BWR, were reviewed against Capability Category I of the ASME PRA standard. Capability Category I was selected as the basis for review due to the specific uses/applications of the SPAR models. The BWR SPAR model was reviewed against 331 ASME PRA Standard Supporting Requirements; however, based on the Capability Category I level of review and the absence of internal flooding and containment performance (LERF) logic only 216 requirements were determined to be applicable. Based on the review, the BWR SPAR model met 139 of the 216 supporting requirements. The review also generated 200 findings or suggestions. Of these 200 findings and suggestions 142 were findings and 58 were suggestions. The PWR SPAR model was also evaluated against the same 331 ASME PRA Standard Supporting Requirements. Of these requirements only 215 were deemed appropriate for the review (for the same reason as noted for the BWR). The PWR review determined that 125 of the 215 supporting requirements met Capability Category I or greater. The review identified 101 findings or suggestions (76 findings and 25 suggestions). These findings or suggestions were developed to identify areas where SPAR models could be enhanced. A process to prioritize and incorporate the findings/suggestions supporting requirements into the SPAR models is being developed. The prioritization process focuses on those findings that will enhance the accuracy, completeness and usability of the SPAR models.

  8. NGNP Project Regulatory Gap Analysis for Modular HTGRs

    SciTech Connect (OSTI)

    Wayne Moe

    2011-09-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) Project Regulatory Gap Analysis (RGA) for High Temperature Gas-Cooled Reactors (HTGR) was conducted to evaluate existing regulatory requirements and guidance against the design characteristics specific to a generic modular HTGR. This final report presents results and identifies regulatory gaps concerning current Nuclear Regulatory Commission (NRC) licensing requirements that apply to the modular HTGR design concept. This report contains appendices that highlight important HTGR licensing issues that were found during the RGA study. The information contained in this report will be used to further efforts in reconciling HTGR-related gaps in the NRC licensing structure, which has to date largely focused on light water reactor technology.

  9. Codes and standards and other guidance cited in regulatory documents

    SciTech Connect (OSTI)

    Nickolaus, J.R.; Bohlander, K.L.

    1996-08-01T23:59:59.000Z

    As part of the U.S. Nuclear Regulatory Commission (NRC) Standard Review Plan Update and Development Program (SRP-UDP), Pacific Northwest National Laboratory developed a listing of industry consensus codes and standards and other government and industry guidance referred to in regulatory documents. The SRP-UDP has been completed and the SRP-Maintenance Program (SRP-MP) is now maintaining this listing. Besides updating previous information, Revision 3 adds approximately 80 citations. This listing identifies the version of the code or standard cited in the regulatory document, the regulatory document, and the current version of the code or standard. It also provides a summary characterization of the nature of the citation. This listing was developed from electronic searches of the Code of Federal Regulations and the NRC`s Bulletins, Information Notices, Circulars, Enforcement Manual, Generic Letters, Inspection Manual, Policy Statements, Regulatory Guides, Standard Technical Specifications and the Standard Review Plan (NUREG-0800).

  10. Nuclear Regulatory Commission Proceedings: A Guide for Intervenors

    E-Print Network [OSTI]

    Hansell, Dean

    1982-01-01T23:59:59.000Z

    License for Floating Nuclear Power Plants). The requirements207 (1978) (Floating Nuclear Power Plants). 101. Early site

  11. U.S. Nuclear Regulatory Commission Extremely Low Probability of Rupture pilot study : xLPR framework model user's guide.

    SciTech Connect (OSTI)

    Kalinich, Donald A.; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-12-01T23:59:59.000Z

    For the U.S. Nuclear Regulatory Commission (NRC) Extremely Low Probability of Rupture (xLPR) pilot study, Sandia National Laboratories (SNL) was tasked to develop and evaluate a probabilistic framework using a commercial software package for Version 1.0 of the xLPR Code. Version 1.0 of the xLPR code is focused assessing the probability of rupture due to primary water stress corrosion cracking in dissimilar metal welds in pressurizer surge nozzles. Future versions of this framework will expand the capabilities to other cracking mechanisms, and other piping systems for both pressurized water reactors and boiling water reactors. The goal of the pilot study project is to plan the xLPR framework transition from Version 1.0 to Version 2.0; hence the initial Version 1.0 framework and code development will be used to define the requirements for Version 2.0. The software documented in this report has been developed and tested solely for this purpose. This framework and demonstration problem will be used to evaluate the commercial software's capabilities and applicability for use in creating the final version of the xLPR framework. This report details the design, system requirements, and the steps necessary to use the commercial-code based xLPR framework developed by SNL.

  12. Indexes to Nuclear Regulatory Commission issuances, July--December 1995

    SciTech Connect (OSTI)

    NONE

    1996-04-01T23:59:59.000Z

    Digests and indexes for issuances of the NRC, the Atomic Safety and Licensing Board, the Administrative Law Judges, the Directors` Decisions, and the Decisions on Petitions for Rulemaking are presented in this document. These digests and indexes are intended to serve as a guide to the issuances. Information elements common to the cases heard and ruled upon are: (1) case name, (2) full text reference, (3) issuance number, (4) issued raised by appellants, (5) legal citations, (6) name of facility and Docket number, (7) subject matter, (8) type of hearing, and (9) type of issuance.

  13. Nuclear Regulatory Commission's Integrated Strategy for Spent Fuel Management

    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's Integrated Strategy

  14. The Regulatory Challenges of Decommissioning Nuclear Power Plants in Korea - 13101

    SciTech Connect (OSTI)

    Lee, Jungjoon; Ahn, Sangmyeon; Choi, Kyungwoo [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)] [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Kim, Juyoul; Kim, Juyub [FNC Technology, 46 Tabsil-ro, Giheung-gu, Yongin 446-902 (Korea, Republic of)] [FNC Technology, 46 Tabsil-ro, Giheung-gu, Yongin 446-902 (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    As of 2012, 23 units of nuclear power plants are in operation, but there is no experience of permanent shutdown and decommissioning of nuclear power plant in Korea. It is realized that, since late 1990's, improvement of the regulatory framework for decommissioning has been emphasized constantly from the point of view of International Atomic Energy Agency (IAEA)'s safety standards. And it is known that now IAEA prepare the safety requirement on decommissioning of facilities, its title is the Safe Decommissioning of Facilities, General Safety Requirement Part 6. According to the result of IAEA's Integrated Regulatory Review Service (IRRS) mission to Korea in 2011, it was recommended that the regulatory framework for decommissioning should require decommissioning plans for nuclear installations to be constructed and operated and these plans should be updated periodically. In addition, after the Fukushima nuclear disaster in Japan in March of 2011, preparedness for early decommissioning caused by an unexpected severe accident became also important issues and concerns. In this respect, it is acknowledged that the regulatory framework for decommissioning of nuclear facilities in Korea need to be improved. First of all, we identify the current status and relevant issues of regulatory framework for decommissioning of nuclear power plants compared to the IAEA's safety standards in order to achieve our goal. And then the plan is to be established for improvement of regulatory framework for decommissioning of nuclear power plants in Korea. After dealing with it, it is expected that the revised regulatory framework for decommissioning could enhance the safety regime on the decommissioning of nuclear power plants in Korea in light of international standards. (authors)

  15. 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.

  16. A compilation of reports of the Advisory Committee on nuclear waste, July 1996--June 1997

    SciTech Connect (OSTI)

    NONE

    1997-08-01T23:59:59.000Z

    This compilation contains 11 reports issued by the Advisory Committee on Nuclear Waste (ACNW) during the ninth year of its operation. The reports were submitted to the Chairman and Commissioners of the U.S. Nuclear Regulatory Commission. All reports prepared by the Committee have been made available to the public through the NRC Public Document Room, the U.S. Library of Congress, and the internet at http://www.nrc.gov/ACRSACNW.

  17. PHYSICAL FIDELITY CONSIDERATIONS FOR NRC ADVANCED REACTOR CONTROL ROOM TRAINING SIMULATORS USED FOR INSPECTOR/EXAMINER TRAINING

    SciTech Connect (OSTI)

    Branch, Kristi M.; Mitchell, Mark R.; Miller, Mark; Cochrum, Steven

    2010-11-07T23:59:59.000Z

    This paper describes research into the physical fidelity requirements of control room simulators to train U.S. Nuclear Regulatory Commission (NRC) staff for their duties as inspectors and license examiners for next-generation nuclear power plants. The control rooms of these power plants are expected to utilize digital instrumentation and controls to a much greater extent than do current plants. The NRC is assessing training facility needs, particularly for control room simulators, which play a central role in NRC training. Simulator fidelity affects both training effectiveness and cost. Research has shown high simulation fidelity sometimes positively affects transfer to the operational environment but sometimes makes no significant difference or actually impedes learning. The conditions in which these different effects occur are often unclear, especially for regulators (as opposed to operators) about whom research is particularly sparse. This project developed an inventory of the tasks and knowledges, skills, and abilities that NRC regulators need to fulfill job duties and used expert panels to characterize the inventory items by type and level of cognitive/behavioral capability needed, difficulty to perform, importance to safety, frequency of performance, and the importance of simulator training for learning these capabilities. A survey of current NRC staff provides information about the physical fidelity of the simulator on which the student trained to the control room to which the student was assigned and the effect lack of fidelity had on learning and job performance. The study concludes that a high level of physical fidelity is not required for effective training of NRC staff.

  18. Nuclear Material Control and Accountability

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

    2011-06-27T23:59:59.000Z

    The Order establishes performance objectives, metrics, and requirements for developing, implementing, and maintaining a nuclear material control and accountability (MC&A) program within the U.S. Department of Energy (DOE), including the National Nuclear Security Administration (NNSA), and for DOE owned materials at other facilities that are exempt from licensing by the Nuclear Regulatory Commission (NRC). Admin Chg 3, dated 5-15-15 cancels Admin Chg 2.

  19. NRC Reviewer Aid for Evaluating the Human Factors Engineering Aspects of Small Modular Reactors

    SciTech Connect (OSTI)

    OHara J. M.; Higgins, J.C.

    2012-01-13T23:59:59.000Z

    Small modular reactors (SMRs) are a promising approach to meeting future energy needs. Although the electrical output of an individual SMR is relatively small compared to that of typical commercial nuclear plants, they can be grouped to produce as much energy as a utility demands. Furthermore, SMRs can be used for other purposes, such as producing hydrogen and generating process heat. The design characteristics of many SMRs differ from those of current conventional plants and may require a distinct concept of operations (ConOps). The U.S. Nuclear Regulatory Commission (NRC) conducted research to examine the human factors engineering (HFE) and the operational aspects of SMRs. The research identified thirty potential human-performance issues that should be considered in the NRC's reviews of SMR designs and in future research activities. The purpose of this report is to support NRC HFE reviewers of SMR applications by identifying some of the questions that can be asked of applicants whose designs have characteristics identified in the issues. The questions for each issue were identified and organized based on the review elements and guidance contained in Chapter 18 of the Standard Review Plan (NUREG-0800), and the Human Factors Engineering Program Review Model (NUREG-0711).

  20. Control and Accountability of Nuclear Materials

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

    2000-11-20T23:59:59.000Z

    To prescribe Department of Energy (DOE) requirements, including those for the National Nuclear Security Administration (NNSA), for nuclear material control and accountability (MC&A) for DOE-owned and -leased facilities and DOE-owned nuclear materials at other facilities that are exempt from licensing by the Nuclear Regulatory Commission (NRC). DOE N 251.60, dated 11-19-04, extends this directive until 11-19-05. Cancels DOE O 474.1.

  1. Regulatory Oversight Program, July 1, 1993--March 3, 1997. Volume 3

    SciTech Connect (OSTI)

    NONE

    1997-12-31T23:59:59.000Z

    On July, 1993, a Regulatory Oversight (RO) organization was established within the US DOE, Oak Ridge Operations (ORO) to provide regulatory oversight of the DOE uranium enrichment facilities leased to the United States Enrichment Corporation (USEC). The purpose of the OR program was to ensure continued plant safety, safeguards and security while the plants were transitioned to regulatory oversight by the Nuclear Regulatory Commission (NRC). Volume 3 contains copies of two reports that document the DOE/ORO regulatory oversight inspection and enforcement history for each gaseous diffusion plant site. Each report provides a formal mechanism by which DOE/ORO could communicate the inspection and enforcement history to NRC. The reports encompass the inspection activities that occurred during July 1, 1993 through March 2, 1997.

  2. 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.

  3. Nuclear Regulatory Commission issuances, Volume 44, No. 4

    SciTech Connect (OSTI)

    NONE

    1996-10-01T23:59:59.000Z

    This report includes the issuances received in October 1996. Issuances are from the Commission, the Atomic Safety and Licensing Boards, and the Directors` Decisions. 15 issuances were received and are abstracted individually in the database: Louisiana Energy Services, U.S. Enrichment Corporation, Yankee Atomic Electric Company, General Public Utilities Nuclear Corporation, James L. Shelton, Juan Guzman, Northern States Power Company, TESTCO Inc., Washington Public Power Supply System, all nuclear plants, Cleveland Electric Illuminating Company, Duke Power Company, Florida Power Corporation, and Northeast Nuclear Energy Company (2 issuances). No issuances were received from the the Administrative Law Judges or the Decisions on Petitions for Rulemaking.

  4. Department of Energy Commends the Nuclear Regulatory Commission...

    Energy Savers [EERE]

    of Energy is proud to foster an environment where nuclear power - a safe and emissions-free source of energy - can begin to thrive," Secretary of Energy Samuel W. Bodman said....

  5. Regulatory Oversight Program, July 1, 1993--March 3, 1997. Volume 1

    SciTech Connect (OSTI)

    NONE

    1997-09-01T23:59:59.000Z

    On July 1, 1993, a Regulatory Oversight (RO) organization was established within the United States Department of Energy (DOE), Oak Ridge Operations (ORO) to provide regulatory oversight of the DOE uranium enrichment facilities leased to the United States Enrichment Corporation (USEC). The purpose of the RO program was to ensure continued plant safety, safeguards and security while the Paducah and Portsmouth gaseous diffusion plants (GDPs) transitioned to regulatory oversight by the Nuclear Regulatory Commission (NRC). These activities were performed under the authority of the lease agreement between DOE and USEC until NRC issued a Certificate of Compliance or approved a Compliance Plan pursuant to Section 1701 of the Atomic Energy Act of 1954, as amended, and assumed regulatory responsibility. This report chronicles the formal development, operation and key activities of the RO organization from its beginning in July 1993, until the turnover of the regulatory oversight responsibility to the NRC on March 3, 1997. Through its evolution to closure, the RO program was a formal, proceduralized effort designed to provide consistent regulation and to facilitate transition to NRC. The RO Program was also a first-of-a-kind program for DOE. The process, experience, and lessons learned summarized herein should be useful as a model for transition of other DOE facilities to privatization or external regulation.

  6. Nuclear Regulatory Commission Issuances. Opinions and decisions of the Nuclear Regulatory Commission with selected orders: July 1, 1992--December 31, 1992, Volume 36, Pages 1--396

    SciTech Connect (OSTI)

    Not Available

    1992-12-31T23:59:59.000Z

    This is the thirty-sixth volume of issuances (1-396) of the Nuclear Regulatory Commission and its Atomic Safety and Licensing Boards, Administrative Law Judges, and Office Directors. It covers the period from July 1, 1992-December 31, 1992. Atomic Safety and Licensing Boards are authorized by Section 191 of the Atomic Energy Act of 1954. These Boards, comprised of three members conduct adjudicatory hearings on applications to construct and operate nuclear power plants and related facilities and issue initial decisions which, subject to internal review and appellate procedures, become the final Commission action with respect to those applications. Boards are drawn from the Atomic Safety and Licensing Board Panel, comprised of lawyers, nuclear physicists and engineers, environmentalists, chemists, and economists. The Atomic Energy Commission first established Licensing Boards in 1962 and the Panel in 1967.

  7. Regulatory cross-cutting topics for fuel cycle facilities.

    SciTech Connect (OSTI)

    Denman, Matthew R.; Brown, Jason; Goldmann, Andrew Scott; Louie, David

    2013-10-01T23:59:59.000Z

    This report overviews crosscutting regulatory topics for nuclear fuel cycle facilities for use in the Fuel Cycle Research&Development Nuclear Fuel Cycle Evaluation and Screening study. In particular, the regulatory infrastructure and analysis capability is assessed for the following topical areas:Fire Regulations (i.e., how applicable are current Nuclear Regulatory Commission (NRC) and/or International Atomic Energy Agency (IAEA) fire regulations to advance fuel cycle facilities)Consequence Assessment (i.e., how applicable are current radionuclide transportation tools to support risk-informed regulations and Level 2 and/or 3 PRA) While not addressed in detail, the following regulatory topic is also discussed:Integrated Security, Safeguard and Safety Requirement (i.e., how applicable are current Nuclear Regulatory Commission (NRC) regulations to future fuel cycle facilities which will likely be required to balance the sometimes conflicting Material Accountability, Security, and Safety requirements.)

  8. Nuclear Regulatory Commission issuances. Volume 44, Number 3

    SciTech Connect (OSTI)

    NONE

    1996-09-01T23:59:59.000Z

    This report includes issuances received during September 1996. After reviewing in detail each of the claims made in this informal proceeding the presiding officer sustained the staff of the USNRC in its determination that the applicant did not pass the written portion of his examination to become a licensed operator of a nuclear power plant. In the proceeding concerning citizen group challenges to the decommissioning plan for the Rowe Yankee power station, the licensing board grants licensee Yankee Atomic Electric Company`s motion for summary disposition.

  9. UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20556

    Office of Scientific and Technical Information (OSTI)

    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 742EnergyOnItem Not FoundInformation DOEInformation Summary Big*Thee n0738F 241.3NUCLEAR

  10. Department of Energy and Nuclear Regulatory Commission Increase Cooperation

    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 China U.S.Contamination ControlDecisionsGeothermalPolicy Actto Advance Global Nuclear

  11. Status of VICTORIA: NRC peer review and recent code applications

    SciTech Connect (OSTI)

    Bixler, N.E. [Sandia National Labs., Albuquerque, NM (United States); Schaperow, J.H. [Nuclear Regulatory Commission, Washington, DC (United States)

    1997-12-01T23:59:59.000Z

    VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS. A summary of the results and recommendations of an independent peer review of VICTORIA by the US Nuclear Regulatory Commission (NRC) is presented, along with recent applications of the code. The latter include analyses of a temperature-induced steam generator tube rupture sequence and post-test analyses of the Phebus FPT-1 test. The next planned Phebus test, FTP-4, will focus on fission product releases from a rubble bed, especially those of the less-volatile elements, and on the speciation of the released elements. Pretest analyses using VICTORIA to estimate the magnitude and timing of releases are presented. The predicted release of uranium is a matter of particular importance because of concern about filter plugging during the test.

  12. NRC review of Electric Power Research Institute's Advanced Light Reactor Utility Requirements Document - Program summary, Project No. 669

    SciTech Connect (OSTI)

    Not Available

    1992-08-01T23:59:59.000Z

    The staff of the US Nuclear Regulatory Commission has prepared Volume 1 of a safety evaluation report (SER), NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document -- Program Summary,'' to document the results of its review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document.'' This SER provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  13. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants (Cooperative Agreement DE-FC03-99SF21902, Am. M004) Final Technical Report

    SciTech Connect (OSTI)

    Stanley E. Ritterbusch, et. al.

    2003-01-29T23:59:59.000Z

    OAK-B135 Research under this project addresses the barriers to long term use of nuclear-generated electricity in the United States. It was agreed that a very basic and significant change to the current method of design and regulation was needed. That is, it was believed that the cost reduction goal could not be met by fixing the current system (i.e., an evolutionary approach) and a new, more advanced approach for this project would be needed. It is believed that a completely new design and regulatory process would have to be developed--a ''clean sheet of paper'' approach. This new approach would start with risk-based methods, would establish probabilistic design criteria, and would implement defense-in-depth only when necessary (1) to meet public policy issues (e.g., use of a containment building no matter how low the probability of a large release is) and (2) to address uncertainties in probabilistic methods and equipment performance. This new approach is significantly different from the Nuclear Regulatory Commission's (NRC) current risk-informed program for operating plants. For our new approach, risk-based methods are the primary means for assuring plant safety, whereas in the NRC's current approach, defense-in-depth remains the primary means of assuring safety. The primary accomplishments in the first year--Phase 1 were (1) the establishment of a new, highly risk-informed design and regulatory framework, (2) the establishment of the preliminary version of the new, highly risk-informed design process, (3) core damage frequency predictions showing that, based on new, lower pipe rupture probabilities, the design of the emergency core cooling system equipment can be simplified without reducing plant safety, and (4) the initial development of methods for including uncertainties in a new integrated structures-systems design model. Under the new regulatory framework, options for the use of ''design basis accidents'' were evaluated. It is expected that design basis accidents would be an inherent part of the Probabilistic Safety Assessment for the plant and their evaluation would be probabilistic. Other first year accomplishments include (1) the conversion of an NRC database for cross-referencing NRC criteria and industry codes and standards to Microsoft 2000 software, (2) an assessment of the NRC's hearing process which concluded that the normal cross-examination during public hearings is not actually required by the U.S. Administrative Procedures Act, (3) the identification and listing of reliability data sources, and (4) interfacing with other industry groups (e.g., NEI and IAEA) and NRC at workshops for risk-informing regulations. The major accomplishments during the second year consisted of (1) issuance of the final report for Subtask 1.1, ''Identify Current Applicable Regulatory Requirements [and Industry Standards],'' (2) issuance of the final report for Subtask 1.2,'' Identify Structures, Systems, and Components and Their Associate d Costs for a Typical Plant,'' (3) extension of the new, highly risk-informed design and regulatory framework to non-light-water-reactor technology, (4) completion of more detailed thermal-hydraulic and probabilistic analyses of advanced conceptual reactor system/component designs, (6) initial evaluation and recommendations for improvement of the NRC design review process, and (7) initial development of the software format, procedures and statistical routines needed to store, analyze and retrieve the available reliability data. Final reports for Subtasks 1.1 (regulatory and design criteria) and 1.2 (costs for structures, systems, and components) were prepared and issued. A final report for Subtask 1.3 (Regulatory Framework) was drafted with the aim to issue it in Phase 3 (Year 3). One technical report was produced for Subtask 1.4 (methods development) and two technical reports were produced for Subtask 1.6 (sample problem analysis). An interim report on the NRC design review process (Subtask 1.7) was prepared and issued. Finally, a report on Subtask 2.2 (database weaknesses) addressed the i

  14. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 2. Evaluation of seismic designs: a review of seismic design requirements for Nuclear Power Plant Piping

    SciTech Connect (OSTI)

    Not Available

    1985-04-01T23:59:59.000Z

    This document reports the position and recommendations of the NRC Piping Review Committee, Task Group on Seismic Design. The Task Group considered overlapping conservation in the various steps of seismic design, the effects of using two levels of earthquake as a design criterion, and current industry practices. Issues such as damping values, spectra modification, multiple response spectra methods, nozzle and support design, design margins, inelastic piping response, and the use of snubbers are addressed. Effects of current regulatory requirements for piping design are evaluated, and recommendations for immediate licensing action, changes in existing requirements, and research programs are presented. Additional background information and suggestions given by consultants are also presented.

  15. Nuclear Regulatory Commission Issuances: Opinions and decisions of the Nuclear Regulatory Commission with selected orders. Progress report, January 1, 1996--June 30, 1996. Volume 43, pages 1-358

    SciTech Connect (OSTI)

    NONE

    1997-05-01T23:59:59.000Z

    The hardbound edition of the Nuclear Regulatory Issuances is a final compilation of the monthly issuances. It includes all legal precedents for the agency within a six month period. This is the forty-third volume of issuances.

  16. Intercomparison run for uranium and tritium determination in urine samples, organised by Nuclear Regulatory Authority, Argentina

    E-Print Network [OSTI]

    Serdeiro, N H; Equillor, H E

    2003-01-01T23:59:59.000Z

    The Nuclear Regulatory Authority (ARN), Argentina, has carried out an intercomparison run for tritium and uranium determination in urine, in November 2002. The aim of this exercise was to assess the performance of the laboratories that usually inform these radionuclides and to provide technical support in order to have an appropriate occupational monitoring in vitro. In the present work, the results of the intercomparison and the assessment of each laboratory are published.

  17. Evaluation of offsite dose assessment methodologies used in the nuclear industry 

    E-Print Network [OSTI]

    Berry, Robert Orwell

    1992-01-01T23:59:59.000Z

    OF CPSES . . . . . IV EXPOSURE PATHWAYS TO MAN Page VI V I I IX 15 21 V DESCRIPTION OF DOSE ASSESSMENT METHODS . . . . 28 NCRP COMMENTARY No. 3 COMPLY AIRDOS-PC ODCM . REGULATORY GUIDE 1. 109 . . . VI COMPARISONS OF DOSE CALCULATION RESULTS...) Commentary No. 3, the Offsite Dose Calculation Manual (ODCM) from CPSES and the Nuclear Regulatory Commission's (NRC) Regulatory Guide 1. 109 were utilized as dose assessment tools during this study The amount of site-specific information entered...

  18. Independent Verification and Validation Of SAPHIRE 8 Software Project Plan Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Carl Wharton

    2009-10-01T23:59:59.000Z

    This report provides an evaluation of the Project Plan. The Project Plan is intended to provide the high-level direction that documents the required software activities to meet the contractual commitments prepared by the sponsor; the Nuclear Regulatory Commission.

  19. Independent Verification and Validation Of SAPHIRE 8 Software Project Plan Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Carl Wharton; Kent Norris

    2010-03-01T23:59:59.000Z

    This report provides an evaluation of the Project Plan. The Project Plan is intended to provide the high-level direction that documents the required software activities to meet the contractual commitments prepared by the sponsor; the Nuclear Regulatory Commission.

  20. Independent Verification and Validation Of SAPHIRE 8 Software Project Plan Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Carl Wharton; Kent Norris

    2009-12-01T23:59:59.000Z

    This report provides an evaluation of the Project Plan. The Project Plan is intended to provide the high-level direction that documents the required software activities to meet the contractual commitments prepared by the sponsor; the Nuclear Regulatory Commission.

  1. Regulatory instrument review: Aging management of LWR cables, containment and basemat, reactor coolant pumps, and motor-operated valves

    SciTech Connect (OSTI)

    Werry, E.V.; Somasundaram, S.

    1995-09-01T23:59:59.000Z

    The results of Stage 2 of the Regulatory Instrument Review are presented in this volume. Selected regulatory instruments, such as the Code of Federal Regulations (CFR), US Nuclear Regulatory Commission (NRC), Regulatory Guides, and ASME Codes, were investigated to determine the extent to which these regulations apply aging management to selected safety-related components in nuclear power plants. The Regulatory Instrument Review was funded by the NRC under the Nuclear Plant Aging Research (NPAR) program. Stage 2 of the review focused on four safety-related structures and components; namely, cables, containment and basemat, reactor coolant pumps, and motor-operated valves. The review suggests that the primary-emphasis of the regulatory instruments was on the design, construction, start-up, and operation of a nuclear power plant, and that aging issues were primarily addressed after an aging-related problem was recognized. This Stage 2 review confirms the results of the prior review; (see Regulatory Instrument Review: Management of Aging of LWR Major Safety-Related Components NUREG/CR-5490. The observations indicate that the regulations generally address management of age-related degradation indirectly. Specific age-related degradation phenomena frequently are dealt with in bulletins and notices or through generic issues, letters, etc. The major recommendation of this report, therefore, is that the regulatory instruments should more directly and explicitly address the aging phenomenon and the management of the age-related degradation process.

  2. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    SciTech Connect (OSTI)

    Charak, I.; Kier, P.H. [Argonne National Lab., IL (United States)

    1995-04-01T23:59:59.000Z

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule ({section}50.62); (2) station blackout ({section}50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term ({section}50.34(f) and {section}100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements ({section}50.55a, etc); (6) ECCS acceptance criteria ({section}50.46)(b); (7) combustible gas control ({section}50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle ({section}51.51); and (11) (standards {section}50.55a).

  3. U.S. Nuclear Regulatory Commission organization charts and functional statements. Revision 18

    SciTech Connect (OSTI)

    NONE

    1995-07-23T23:59:59.000Z

    This document (NUREG-0325) is the current US NRC organization chart, listing all NRC offices and regions and their components down through the branch level as of July 23, 1995. Functional statements of each position are given, as is the name of the individual holding the position.

  4. Vital area determination techniques at nuclear power plants

    SciTech Connect (OSTI)

    Pan, P.Y.

    1987-07-01T23:59:59.000Z

    This paper describes the vital area determination programs being conducted at the Los Alamos National Laboratory to support the Nuclear Regulatory Commission (NRC) in evaluating nuclear power plant licensees' compliance with safeguards/security requirements. These projects, the Vital Area Analysis (VAA) Program and the Vital Equipment Determination Techniques Research Study (VEDTRS), are designed to identify a plant's vital areas and to develop protection strategies against adversary threats in nuclear power plants.

  5. NRC Construction Light Source Flicker: What We

    E-Print Network [OSTI]

    California at Davis, University of

    NRC Construction Light Source Flicker: What We Need to Know, and Why You Should Care NRC Construction Jennifer A. Veitch, Ph.D. (c) 2013, National Research Council Canada #12;NRC Construction Handbook: Reference & Application (9th Ed.), 2000, p. 3-20 #12;NRC Construction Flicker Effects 1

  6. Nuclear Matrix Protein Matrin3 regulates alternative splicing and forms overlapping regulatory networks with PTB

    E-Print Network [OSTI]

    Coelho, Miguel B.; Attig, Jan; Bellora, Nicolás; König, Julian; Hallegger, Martina; Kayikci, Melis; Eyras, Eduardo; Ule, Jernej; Smith, Christopher W.J.

    2015-01-19T23:59:59.000Z

    The EMBO Journal Miguel B Coelho et al Matrin3 regulates splicing The EMBO Journal 3 et al, 2010). The array data were analysed using the ASPIRE3 pipeline. Only 61 genes showed changes in RNA levels of greater than twofold, including the expected reduction... Article Nuclear matrix protein Matrin3 regulates alternative splicing and forms overlapping regulatory networks with PTB Miguel B Coelho1, Jan Attig2,3, Nicolás Bellora4,5,6, Julian König3,†, Martina Hallegger1,2, Melis Kayikci3, Eduardo Eyras4...

  7. Regulatory and Financial Reform of Federal Research Policy: Recommenda...

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

    and Financial Reform of Federal Research Policy: Recommendations to the NRC Committee on Research Universities Regulatory and Financial Reform of Federal Research Policy:...

  8. EPRI/NRC-RES fire human reliability analysis guidelines.

    SciTech Connect (OSTI)

    Lewis, Stuart R. (Electric Power Research Institute, Charlotte, NC); Cooper, Susan E. (U.S. Nuclear Regulatory Commission, Rockville, MD); Najafi, Bijan (SAIC, Campbell, CA); Collins, Erin (SAIC, Campbell, CA); Hannaman, Bill (SAIC, Campbell, CA); Kohlhepp, Kaydee (Scientech, Tukwila, WA); Grobbelaar, Jan (Scientech, Tukwila, WA); Hill, Kendra (U.S. Nuclear Regulatory Commission, Rockville, MD); Hendrickson, Stacey M. Langfitt; Forester, John Alan; Julius, Jeff (Scientech, Tukwila, WA)

    2010-03-01T23:59:59.000Z

    During the 1990s, the Electric Power Research Institute (EPRI) developed methods for fire risk analysis to support its utility members in the preparation of responses to Generic Letter 88-20, Supplement 4, 'Individual Plant Examination - External Events' (IPEEE). This effort produced a Fire Risk Assessment methodology for operations at power that was used by the majority of U.S. nuclear power plants (NPPs) in support of the IPEEE program and several NPPs overseas. Although these methods were acceptable for accomplishing the objectives of the IPEEE, EPRI and the U.S. Nuclear Regulatory Commission (NRC) recognized that they required upgrades to support current requirements for risk-informed, performance-based (RI/PB) applications. In 2001, EPRI and the USNRC's Office of Nuclear Regulatory Research (RES) embarked on a cooperative project to improve the state-of-the-art in fire risk assessment to support a new risk-informed environment in fire protection. This project produced a consensus document, NUREG/CR-6850 (EPRI 1011989), entitled 'Fire PRA Methodology for Nuclear Power Facilities' which addressed fire risk for at power operations. NUREG/CR-6850 developed high level guidance on the process for identification and inclusion of human failure events (HFEs) into the fire PRA (FPRA), and a methodology for assigning quantitative screening values to these HFEs. It outlined the initial considerations of performance shaping factors (PSFs) and related fire effects that may need to be addressed in developing best-estimate human error probabilities (HEPs). However, NUREG/CR-6850 did not describe a methodology to develop best-estimate HEPs given the PSFs and the fire-related effects. In 2007, EPRI and RES embarked on another cooperative project to develop explicit guidance for estimating HEPs for human failure events under fire generated conditions, building upon existing human reliability analysis (HRA) methods. This document provides a methodology and guidance for conducting a fire HRA. This process includes identification and definition of post-fire human failure events, qualitative analysis, quantification, recovery, dependency, and uncertainty. This document provides three approaches to quantification: screening, scoping, and detailed HRA. Screening is based on the guidance in NUREG/CR-6850, with some additional guidance for scenarios with long time windows. Scoping is a new approach to quantification developed specifically to support the iterative nature of fire PRA quantification. Scoping is intended to provide less conservative HEPs than screening, but requires fewer resources than a detailed HRA analysis. For detailed HRA quantification, guidance has been developed on how to apply existing methods to assess post-fire fire HEPs.

  9. Review of maintenance personnel practices at nuclear power plants

    SciTech Connect (OSTI)

    Chockie, A.D.; Badalamente, R.V.; Hostick, C.J.; Vickroy, S.C.; Bryant, J.L.; Imhoff, C.H.

    1984-05-01T23:59:59.000Z

    As part of the Nuclear Regulatory Commission (NRC) sponsored Maintenance Qualifications and Staffing Project, the Pacific Northwest Laboratory (PNL) has conducted a preliminary assessment of nuclear power plant (NPP) maintenance practices. As requested by the NRC, the following areas within the maintenance function were examined: personnel qualifications, maintenance training, overtime, shiftwork and staffing levels. The purpose of the assessment was to identify the primary safety-related problems that required further analysis before specific recommendations can be made on the regulations affecting NPP maintenance operations.

  10. Seismic risk assessment as applied to the Zion Nuclear Generating Station

    SciTech Connect (OSTI)

    Wells, J.

    1984-08-01T23:59:59.000Z

    To assist the US Nuclear Regulatory Commission (NRC) in its licensing and evaluation role, the NRC funded the Seismic Safety Margins Research Program (SSMRP) at Lawrence Livermore National Laboratory (LLNL) with the goal of developing tools and data bases to evaluate the risk of earthquake caused radioactive release from a commercial nuclear power plant. This paper describes the SSMRP risk assessment methodology and the results generated by applying this methodology to the Zion Nuclear Generating Station. In addition to describing the failure probabilities and risk values, the effects of assumptions about plant configuration, plant operation, and dependence will be given.

  11. From 1998 to 2000, through the Interagency Steering Committee on Radiation Standards (ISCORS), the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Environmental

    E-Print Network [OSTI]

    . The parameters used in the dose modeling for the scenarios and the dose conversion factors calculated for each were taken. Approximately half of the samples were analyzed by the U.S. Department of Energy's Oak

  12. Fitness for duty in the nuclear industry: Update of the technical issues 1996

    SciTech Connect (OSTI)

    Durbin, N.; Grant, T. [eds.] [Battelle Seattle Research Center, WA (United States)

    1996-05-01T23:59:59.000Z

    The purpose of this report is to provide an update of information on the technical issues surrounding the creation, implementation, and maintenance of fitness-for-duty (FFD) policies and programs. It has been prepared as a resource for Nuclear Regulatory Commission (NRC) and nuclear power plant personnel who deal with FFD programs. It contains a general overview and update on the technical issues that the NRC considered prior to the publication of its original FFD rule and the revisions to that rule (presented in earlier NUREG/CRs). It also includes chapters that address issues about which there is growing concern and/or about which there have been substantial changes since NUREG/CR-5784 was published. Although this report is intended to support the NRC`s rule making on fitness for duty, the conclusions of the authors of this report are their own and do not necessarily represent the opinions of the NRC.

  13. Report to the NRC on guidance for preparing scenarios for emergency preparedness exercises at nuclear generating stations. Draft report for comment

    SciTech Connect (OSTI)

    Martin, G.F.; Hickey, E.E.; Moeller, M.P.; Schultz, D.H.; Bethke, G.W.

    1986-03-01T23:59:59.000Z

    A scenario guidance handbook was prepared to assist emergency planners in developing scenarios for emergency preparedness exercises at nuclear power plants. The handbook provides guidance for the development of the objectives of an exercise, the descriptions of scenario events and responses, and the instructions to the participants. Information concerning implementation of the scenario, critiques and findings, and generation and format of scenario data are also included. Finally, examples of manual calculational techniques for producing radiological data are included as an appendix.

  14. LBB evaluation for a typical Japanese PWR primary loop by using the US NRC approved methods

    SciTech Connect (OSTI)

    Swamy, S.A.; Bhowmick, D.C.; Prager, D.E. [Westinghouse Nuclear Technology Division, Pittsburgh, PA (United States)

    1997-04-01T23:59:59.000Z

    The regulatory requirements for postulated pipe ruptures have changed significantly since the first nuclear plants were designed. The Leak-Before-Break (LBB) methodology is now accepted as a technically justifiable approach for eliminating postulation of double-ended guillotine breaks (DEGB) in high energy piping systems. The previous pipe rupture design requirements for nuclear power plant applications are responsible for all the numerous and massive pipe whip restraints and jet shields installed for each plant. This results in significant plant congestion, increased labor costs and radiation dosage for normal maintenance and inspection. Also the restraints increase the probability of interference between the piping and supporting structures during plant heatup, thereby potentially impacting overall plant reliability. The LBB approach to eliminate postulating ruptures in high energy piping systems is a significant improvement to former regulatory methodologies, and therefore, the LBB approach to design is gaining worldwide acceptance. However, the methods and criteria for LBB evaluation depend upon the policy of individual country and significant effort continues towards accomplishing uniformity on a global basis. In this paper the historical development of the U.S. LBB criteria will be traced and the results of an LBB evaluation for a typical Japanese PWR primary loop applying U.S. NRC approved methods will be presented. In addition, another approach using the Japanese LBB criteria will be shown and compared with the U.S. criteria. The comparison will be highlighted in this paper with detailed discussion.

  15. Independent Verification and Validation Of SAPHIRE 8 Risk Management Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Kent Norris

    2009-11-01T23:59:59.000Z

    This report provides an evaluation of the risk management. Risk management is intended to ensure a methodology for conducting risk management planning, identification, analysis, responses, and monitoring and control activities associated with the SAPHIRE project work, and to meet the contractual commitments prepared by the sponsor; the Nuclear Regulatory Commission.

  16. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

    SciTech Connect (OSTI)

    Whitehead, Donnie Wayne; Varnado, G. Bruce

    2008-09-01T23:59:59.000Z

    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilistic risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.

  17. Scenario guidance handbook for emergency-preparedness exercises at nuclear facilities

    SciTech Connect (OSTI)

    Laughlin, G.J.; Martin, G.F.; Desrosiers, A.E.

    1983-01-01T23:59:59.000Z

    As part of the Emergency Preparedness Implementation Appraisal Program conducted by the Nuclear Regulatory Commission (NRC) with the technical assistance of the Pacific Northwest Laboratory (PNL), emergency preparedness exercises are observed on an annual basis at all licensed reactor facilities. One of the significant findings to arise from these observations was that a large number of the commonly observed problems originated in the design of the scenarios used as a basis for each exercise. In an effort to help eliminate some of these problems a scenario guidance handbook has been generated by PNL for the NRC to assist nuclear power plant licensees in developing scenarios for emergency preparedness exercises.

  18. Greater-than-Class C low-level radioactive waste characterization. Appendix E-5: Impact of the 1993 NRC draft Branch Technical Position on concentration averaging of greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Tuite, P.; Tuite, K.; Harris, G. [Waste Management Group, Inc., Peekskill, NY (United States)

    1994-09-01T23:59:59.000Z

    This report evaluates the effects of concentration averaging practices on the disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) generated by the nuclear utility industry and sealed sources. Using estimates of the number of waste components that individually exceed Class C limits, this report calculates the proportion that would be classified as GTCC LLW after applying concentration averaging; this proportion is called the concentration averaging factor. The report uses the guidance outlined in the 1993 Nuclear Regulatory Commission (NRC) draft Branch Technical Position on concentration averaging, as well as waste disposal experience at nuclear utilities, to calculate the concentration averaging factors for nuclear utility wastes. The report uses the 1993 NRC draft Branch Technical Position and the criteria from the Barnwell, South Carolina, LLW disposal site to calculate concentration averaging factors for sealed sources. The report addresses three waste groups: activated metals from light water reactors, process wastes from light-water reactors, and sealed sources. For each waste group, three concentration averaging cases are considered: high, base, and low. The base case, which is the most likely case to occur, assumes using the specific guidance given in the 1993 NRC draft Branch Technical Position on concentration averaging. To project future GTCC LLW generation, each waste category is assigned a concentration averaging factor for the high, base, and low cases.

  19. A compilation of reports of the Advisory Committee on Reactor Safeguards, 1997 annual, U.S. Nuclear Regulatory Commission. Volume 19

    SciTech Connect (OSTI)

    NONE

    1998-04-01T23:59:59.000Z

    This compilation contains 67 ACRS reports submitted to the Commission, or to the Executive Director for Operations, during calendar year 1997. It also includes a report to the Congress on the NRC Safety Research Program. Specific topics include: (1) advanced reactor designs, (2) emergency core cooling systems, (3) fire protection, (4) generic letters and issues, (5) human factors, (6) instrumentation, control and protection systems, (7) materials engineering, (8) probabilistic risk assessment, (9) regulatory guides and procedures, rules, regulations, and (10) safety research, philosophy, technology and criteria.

  20. Final report of the NRC-Agreement State Working Group to evaluate control and accountability of licensed devices

    SciTech Connect (OSTI)

    NONE

    1996-10-01T23:59:59.000Z

    US NRC staff acknowledged that licensees were having problems maintaining control over and accountability for devices containing radioactive material. In June 1995, NRC approved the staff`s suggestion to form a joint NRC-Agreement State Working Group to evaluate the problem and propose solutions. The staff indicated that the Working Group was necessary to address the concerns from a national perspective, allow for a broad level of Agreement State input, and to reflect their experience. Agreement State participation in the process was essential since some Agreement States have implemented effective programs for oversight of device users. This report includes the 5 recommendations proposed by the Working Group to increase regulatory oversight, increase control and accountability of devices, ensure proper disposal, and ensure disposal of orphaned devices. Specifically, the Working Group recommends that: (1) NRC and Agreement States increase regulatory oversight for users of certain devices; (2) NRC and Agreement State impose penalties on persons losing devices; (3) NRC and Agreement States ensure proper disposal of orphaned devices; (4) NRC encourage States to implement similar oversight programs for users of Naturally-Occurring or Accelerator- Produced Material; and (5) NRC encourage non-licensed stakeholders to take appropriate actions, such as instituting programs for material identification.

  1. Site Selection & Characterization Status Report for Next Generation Nuclear Plant (NGNP)

    SciTech Connect (OSTI)

    Mark Holbrook

    2007-09-01T23:59:59.000Z

    In the near future, the US Department of Energy (DOE) will need to make important decisions regarding design and construction of the Next Generation Nuclear Plant (NGNP). One part of making these decisions is considering the potential environmental impacts that this facility may have, if constructed here at the Idaho National Laboratory (INL). The National Environmental Policy Act (NEPA) of 1969 provides DOE decision makers with a process to systematically consider potential environmental consequences of agency decisions. In addition, the Energy Policy Act of 2005 (Title VI, Subtitel C, Section 644) states that the 'Nuclear Regulatory Commission (NRC) shall have licensing and regulatory authority for any reactor authorized under this subtitle.' This stipulates that the NRC will license the NGNP for operation. The NRC NEPA Regulations (10 CFR Part 51) require tha thte NRC prepare an Environmental Impact Statement (EIS) for a permit to construct a nuclear power plant. The applicant is required to submit an Environmental report (ER) to aid the NRC in complying with NEPA.

  2. Directory of certificiates of compliance for radioactive materials packages: Report of NRC approved packages. Revision 19, Volume 1

    SciTech Connect (OSTI)

    NONE

    1996-10-01T23:59:59.000Z

    This directory provides information on packagings approved by the U.S. Nuclear Regulatory Commission.

  3. 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.

  4. 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.

  5. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 4. Evaluation of other loads and load combinations

    SciTech Connect (OSTI)

    Not Available

    1984-12-01T23:59:59.000Z

    Six topical areas were covered by the Task Group on Other Dynamic Loads and Load Combinations as described below: Event Combinations - dealing with the potential simultaneous occurrence of earthquakes, pipe ruptures, and water hammer events in the piping design basis; Response Combinations - dealing with multiply supported piping with independent inputs, the sequence of combinations between spacial and modal components of response, and the treatment of high frequency modes in combination with low frequency modal responses; Stress Limits/Dynamic Allowables - dealing with inelastic allowables for piping and strain rate effects; Water Hammer Loadings - dealing with code and design specifications for these loadings and procedures for identifying potential water hammer that could affect safety; Relief Valve Opening and Closing Loads - dealing with the adequacy of analytical tools for predicting the effects of these events and, in addition, with estimating effective cycles for fatigue evaluations; and Piping Vibration Loads - dealing with evaluation procedures for estimating other than seismic vibratory loads, the need to consider reciprocating and rotary equipment vibratory loads, and high frequency vibratory loads. NRC staff recommendations or regulatory changes and additional study appear in this report.

  6. Knowledge Management Initiatives Used to Maintain Regulatory Expertise in Transportation and Storage of Radioactive Materials - 12177

    SciTech Connect (OSTI)

    Lindsay, Haile; Garcia-Santos, Norma; Saverot, Pierre; Day, Neil; Gambone Rodriguez, Kimberly; Cruz, Luis; Sotomayor-Rivera, Alexis; Vechioli, Lucieann; Vera, John; Pstrak, David [United States Nuclear Regulatory Commission, Mail Stop EBB-03D-02M, 6003 Executive Boulevard, Rockville, MD 20852 (United States)

    2012-07-01T23:59:59.000Z

    The U.S. Nuclear Regulatory Commission (NRC) was established in 1974 with the mission to license and regulate the civilian use of nuclear materials for commercial, industrial, academic, and medical uses in order to protect public health and safety, and the environment, and promote the common defense and security. Currently, approximately half (?49%) of the workforce at the NRC has been with the Agency for less than six years. As part of the Agency's mission, the NRC has partial responsibility for the oversight of the transportation and storage of radioactive materials. The NRC has experienced a significant level of expertise leaving the Agency due to staff attrition. Factors that contribute to this attrition include retirement of the experienced nuclear workforce and mobility of staff within or outside the Agency. Several knowledge management (KM) initiatives have been implemented within the Agency, with one of them including the formation of a Division of Spent Fuel Storage and Transportation (SFST) KM team. The team, which was formed in the fall of 2008, facilitates capturing, transferring, and documenting regulatory knowledge for staff to effectively perform their safety oversight of transportation and storage of radioactive materials, regulated under Title 10 of the Code of Federal Regulations (10 CFR) Part 71 and Part 72. In terms of KM, the SFST goal is to share critical information among the staff to reduce the impact from staff's mobility and attrition. KM strategies in place to achieve this goal are: (1) development of communities of practice (CoP) (SFST Qualification Journal and the Packaging and Storing Radioactive Material) in the on-line NRC Knowledge Center (NKC); (2) implementation of a SFST seminar program where the seminars are recorded and placed in the Agency's repository, Agency-wide Documents Access and Management System (ADAMS); (3) meeting of technical discipline group programs to share knowledge within specialty areas; (4) development of written guidance to capture 'administrative and technical' knowledge (e.g., office instructions (OIs), generic communications (e.g., bulletins, generic letters, regulatory issue summary), standard review plans (SRPs), interim staff guidance (ISGs)); (5) use of mentoring strategies for experienced staff to train new staff members; (6) use of Microsoft SharePoint portals in capturing, transferring, and documenting knowledge for staff across the Division from Division management and administrative assistants to the project managers, inspectors, and technical reviewers; and (7) development and implementation of a Division KM Plan. A discussion and description of the successes and challenges of implementing these KM strategies at the NRC/SFST will be provided. (authors)

  7. 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.

  8. Aging management guideline for commercial nuclear power plants - tanks and pools

    SciTech Connect (OSTI)

    Blocker, E.; Smith, S.; Philpot, L.; Conley, J.

    1996-02-01T23:59:59.000Z

    Continued operation of nuclear power plants for periods that extend beyond their original 40-year license period is a desirable option for many U.S. utilities. U.S. Nuclear Regulatory Commission (NRC) approval of operating license renewals is necessary before continued operation becomes a reality. Effective aging management for plant components is important to reliability and safety, regardless of current plant age or extended life expectations. However, the NRC requires that aging evaluations be performed and the effectiveness of aging management programs be demonstrated for components considered within the scope of license renewal before granting approval for operation beyond 40 years. Both the NRC and the utility want assurance that plant components will be highly reliable during both the current license term and throughout the extended operating period. In addition, effective aging management must be demonstrated to support Maintenance Rule (10 CFR 50.65) activities.

  9. NRC Leadership Expectations and Practices for Sustaining a High...

    Office of Environmental Management (EM)

    NRC Leadership Expectations and Practices for Sustaining a High Performing Organization NRC Leadership Expectations and Practices for Sustaining a High Performing Organization May...

  10. NRC Consultation and Monitoring at the Savannah River Site: Focusing Reviews of Two Different Disposal Actions - 12181

    SciTech Connect (OSTI)

    Ridge, A. Christianne; Barr, Cynthia S.; Pinkston, Karen E.; Parks, Leah S.; Grossman, Christopher J.; Alexander, George W. [U.S. Nuclear Regulatory Commission (United States)

    2012-07-01T23:59:59.000Z

    Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the U.S. Department of Energy (DOE) to consult with the U.S. Nuclear Regulatory Commission (NRC) for certain non-high level waste determinations. The NDAA also requires NRC to monitor DOE's disposal actions related to those determinations. In Fiscal Year 2011, the NRC staff reviewed DOE performance assessments for tank closure at the F-Tank Farm (FTF) Facility and salt waste disposal at the Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) as part of consultation and monitoring, respectively. Differences in inventories, waste forms, and key barriers led to different areas of focus in the NRC reviews of these two activities at the SRS. Because of the key role of chemically reducing grouts in both applications, the evaluation of chemical barriers was significant to both reviews. However, radionuclide solubility in precipitated metal oxides is expected to play a significant role in FTF performance whereas release of several key radionuclides from the SDF is controlled by sorption or precipitation within the cementitious wasteform itself. Similarly, both reviews included an evaluation of physical barriers to flow, but differences in the physical configurations of the waste led to differences in the reviews. For example, NRC's review of the FTF focused on the modeled degradation of carbon steel tank liners while the staff's review of the SDF performance included a detailed evaluation of the physical degradation of the saltstone wasteform and infiltration-limiting closure cap. Because of the long time periods considered (i.e., tens of thousands of years), the NRC reviews of both facilities included detailed evaluation of the engineered chemical and physical barriers. The NRC staff reviews of residual waste disposal in the FTF and salt waste disposal in the SDF focused on physical barriers to flow and chemical barriers to radionuclide release from the waste. Because the waste inventory and concentration at both sites is sufficient to generate unacceptable doses to an off-site member of the public or inadvertent intruder in the absence of engineered barriers, the NRC staff review focused on the engineering features DOE plans to put in place to limit radionuclide release. At the FTF, DOE expects that peak doses are delayed beyond a 10,000 year performance period by a combination of (1) the flow-limiting effect of the steel tank liner and (2) chemical conditions created by the stabilizing grout overlying the waste that limit the solubility of key radionuclides for tens of thousands of years. At the SDF, DOE expects that flow will be significantly limited by water shedding along the closure cap lower drainage layer and that radionuclide release will be further limited by radionuclide precipitation or sorption within the high pH, chemically reducing conditions created within the saltstone waste form. Because the performance of both facilities depends on the performance of engineered barriers for thousands of years, the reviews included a detailed evaluation of the expected long-term behavior of these barriers. As previously discussed, NRC staff reviews of DOE waste determinations during consultation are designed to evaluate the three NDAA criteria, whereas the review of an updated PA during monitoring only addresses whether the NRC staff has reasonable assurance that the planned disposal action will meet the performance objectives of 10 CFR Part 61. The NRC staff review of the Waste Determination for the FTF did not include conclusions about whether the planned disposal of residual waste at the FTF would meet the NDAA criteria because of the substantial uncertainties in the degree of waste removal DOE would achieve and other technical uncertainties. The main product of the NRC staff review of the planned FTF disposal action is the recommendation that DOE should conduct waste release experiments to increase support for key modeling assumptions related to: (1) the evolution of pH and Eh in the grouted tank syst

  11. Slovak Nuclear Regulatory Body Position in the Transport of Radioactive Waste

    SciTech Connect (OSTI)

    Homola, J.

    2003-02-27T23:59:59.000Z

    This paper describes safety requirements for transport of radioactive waste in Slovakia and the role of regulatory body in the transport licensing and assessment processes. Importance of radioactive waste shipments have been increased since 1999 by starting of NPP A-1 decommissioning and operation of near surface disposal facility. Also some information from history of shipment as well as future activities are given. Legal basis for radioactive waste transport is resulting from IAEA recommendations in this area. Different types of transport equipment were approved by regulatory body for both liquid and solid waste and transportation permits were issued to their shipment. Regulatory body attention during evaluation of transport safety is focused mainly on ability of individual packages to withstand different transport conditions and on safety analyses performed for transport equipment for liquid waste with high frequency of shipments. During past three years no event was occurred in connection with radioactive waste transport in Slovakia.

  12. Technical description of the NRC long-term whole-rod and crud performance test

    SciTech Connect (OSTI)

    Einziger, R.E.; Fish, R.L.; Knecht, R.L.

    1982-09-01T23:59:59.000Z

    Westinghouse Hanford Company (WHC) and EG and G-Idaho are jointly conducting a long-term, low-temperature, spent-fuel, whole rod and crud behavior test to provide the Nuclear Regulatory Commission (NRC) with information to assist in the licensing of light water reactor (LWR) spent-fuel, dry storage facilities. Readily available fuel rods from an H.B. Robinson Unit 2 (PWR) fuel assembly and a Peach Bottom-II (BWR) fuel assembly were selected for use in the 50-month test. Both intact and defected rods will be tested in inert and oxidizing atmospheres. A 230/sup 0/C test temperature was selected for the first 10-month run. Both nondestructive and destructive examinations are planned to characterize the fuel rod behavior during the 5-y test. Four interim examinations and a final examination will be conducted. Crud spallation behavior will be investigated by sampling the crud particulate from the test capsules at each of the four interim examinations and at the end of the test. The background to whole rod testing, description of rod breach mechanisms, and a detailed description of the test are presented in this document.

  13. Recent Updates to NRC Fuel Performance Codes and Plans for Future Improvements

    SciTech Connect (OSTI)

    Geelhood, Kenneth J.

    2011-12-31T23:59:59.000Z

    FRAPCON-3.4a and FRAPTRAN 1.4 are the most recent versions of the U.S. Nuclear Regulatory Commission (NRC) steady-state and transient fuel performance codes, respectively. These codes have been assessed against separate effects data and integral assessment data and have been determined to provide a best estimate calculation of fuel performance. Recent updates included in FRAPCON-3.4a include updated material properties models, models for new fuel and cladding types, cladding finite element analysis capability, and capability to perform uncertainty analyses and calculate upper tolerance limits for important outputs. Recent updates included in FRAPTRAN 1.4 include: material properties models that are consistent with FRAPCON-3.4a, cladding failure models that are applicable for loss-of coolant-accident and reactivity initiated accident modeling, and updated heat transfer models. This paper briefly describes these code updates and data assessments, highlighting the particularly important improvements and data assessments. This paper also discusses areas of improvements that will be addressed in upcoming code versions.

  14. LWRS II&C Industry and Regulatory Engagement Activities for FY 11

    SciTech Connect (OSTI)

    Ken Thomas

    2011-09-01T23:59:59.000Z

    To ensure broad industry support and coordination for the Advanced Instrumentation, Information, and Controls (II&C) Systems Technologies research pathway, an engagement process will be continually pursued with nuclear asset owners, vendors, and suppliers, Nuclear Regulatory Commission (NRC), and the major industry support organizations of Electric Power Research Institute (EPRI), Institute of Nuclear Power Operations (INPO), and Nuclear Energy Institute (NEI). Nuclear asset owner engagement is a necessary and enabling activity to obtain data and accurate characterization of long-term operational challenges, assess the suitability of proposed research for addressing long-term needs, and gain access to data and representative infrastructure and expertise needed to ensure success of the proposed research and development (R&D) activities. Engagement with vendors and suppliers will ensure that vendor expectations and needs can be translated into requirements that can be met through technology commercialization.

  15. Regulatory Closure Options for the Residue in the Hanford Site Single-Shell Tanks

    SciTech Connect (OSTI)

    Cochran, J.R. Shyr, L.J.

    1998-10-05T23:59:59.000Z

    Liquid, mixed, high-level radioactive waste (HLW) has been stored in 149 single-shell tanks (SSTS) located in tank farms on the U.S. Department of Energy's (DOE's) Hanford Site. The DOE is developing technologies to retrieve as much remaining HLW as technically possible prior to physically closing the tank farms. In support of the Hanford Tanks Initiative, Sandia National Laboratories has addressed the requirements for the regulatory closure of the radioactive component of any SST residue that may remain after physical closure. There is significant uncertainty about the end state of each of the 149 SSTS; that is, the nature and amount of wastes remaining in the SSTS after retrieval is uncertain. As a means of proceeding in the face of these uncertainties, this report links possible end-states with associated closure options. Requirements for disposal of HLW and low-level radioactive waste (LLW) are reviewed in detail. Incidental waste, which is radioactive waste produced incidental to the further processing of HLW, is then discussed. If the low activity waste (LAW) fraction from the further processing of HLW is determined to be incidental waste, then DOE can dispose of that incidental waste onsite without a license from the U.S. Nuclear Regulatory Commissions (NRC). The NRC has proposed three Incidental Waste Criteria for determining if a LAW fraction is incidental waste. One of the three Criteria is that the LAW fraction should not exceed the NRC's Class C limits.

  16. NRC staff site characterization analysis of the Department of Energy`s Site Characterization Plan, Yucca Mountain Site, Nevada

    SciTech Connect (OSTI)

    NONE

    1989-08-01T23:59:59.000Z

    This Site Characterization Analysis (SCA) documents the NRC staff`s concerns resulting from its review of the US Department of Energy`s (DOE`s) Site Characterization Plan (SCP) for the Yucca Mountain site in southern Nevada, which is the candidate site selected for characterization as the nation`s first geologic repository for high-level radioactive waste. DOE`s SCP explains how DOE plans to obtain the information necessary to determine the suitability of the Yucca Mountain site for a repository. NRC`s specific objections related to the SCP, and major comments and recommendations on the various parts of DOE`s program, are presented in SCA Section 2, Director`s Comments and Recommendations. Section 3 contains summaries of the NRC staff`s concerns for each specific program, and Section 4 contains NRC staff point papers which set forth in greater detail particular staff concerns regarding DOE`s program. Appendix A presents NRC staff evaluations of those NRC staff Consultation Draft SCP concerns that NRC considers resolved on the basis of the SCP. This SCA fulfills NRC`s responsibilities with respect to DOE`s SCP as specified by the Nuclear Waste Policy Act (NWPA) and 10 CFR 60.18. 192 refs., 2 tabs.

  17. Improved dose assessment in a nuclear reactor accident using the old and new ICRP methodologies

    E-Print Network [OSTI]

    Yoon, Suk-Chul

    1987-01-01T23:59:59.000Z

    Comparison of organ doses: SST-4 SITE-A . . . 89 90 91 25 Comparison of organ doses: BWR-1 SITE-A . . . 26 Comparison of organ doses: BWR-3 SITE-A . . . 92 93 INTRODUCTION The International Commission on Radiological Protection issued ICRP... section. The ICRP 26 and 30 recommendations and internal dosimetry techniques have been reviewed by the National Council on Radiological Protection and Measurements (NCRP) and the Nuclear Regulatory Commission (NRC). The NCRP review has been published...

  18. Nuclear Energy Research Brookhaven National

    E-Print Network [OSTI]

    Ohta, Shigemi

    Nuclear Energy Research Brookhaven National Laboratory William C. Horak, Chair Nuclear Science and Technology Department #12;BNL Nuclear Energy Research Brookhaven Graphite Research Reactor - 1948 National Nuclear Data Center - 1952* High Flux Beam Reactor - 1964 Technical Support for NRC - 1974

  19. Guide to NRC reporting and recordkeeping requirements. Compiled from requirements in Title 10 of the U.S. Code of Federal Regulations as codified on December 31, 1993; Revision 1

    SciTech Connect (OSTI)

    Collins, M.; Shelton, B.

    1994-07-01T23:59:59.000Z

    This compilation includes in the first two sections the reporting and recordkeeping requirements applicable to US Nuclear Regulatory Commission (NRC) licensees and applicants and to members of the public. It includes those requirements codified in Title 10 of the code of Federal Regulations, Chapter 1, on December 31, 1993. It also includes, in a separate section, any of those requirements that were superseded or discontinued between January 1992 and December 1993. Finally, the appendix lists mailing and delivery addresses for NRC Headquarters and Regional Offices mentioned in the compilation. The Office of Information Resources Management staff compiled this listing of reporting and recordkeeping requirements to briefly describe each in a single document primarily to help licensees readily identify the requirements. The compilation is not a substitute for the regulations, and is not intended to impose any new requirements or technical positions. It is part of NRC`s continuing efforts to comply with the Paperwork Reduction Act of 1980 and the Office of Management and Budget regulations that mandate effective and efficient Federal information resources management programs.

  20. Microsoft PowerPoint - Nuclear Material Import Export License...

    National Nuclear Security Administration (NNSA)

    Commission (NRC) importexport license? - Licenses for imports - Licenses for exports - Export license monitoring Highlights of importsexports in 2014 - Nuclear...

  1. Public Comments on the proposed 10 CFR Part 51 rule for renewal of nuclear power plant operating licenses and supporting documents: Review of concerns and NRC staff response. Appendices

    SciTech Connect (OSTI)

    NONE

    1996-05-01T23:59:59.000Z

    This volume contains several appendices. Appendix A contains the list of individuals and organizations providing comments at various stages of the rulemaking process. The names of commenters at the public meetings are listed in the order that they spoke at the meeting; those who submitted written comments are listed by docket number. Appendix B contains the summaries of comments made. Each comment summary is identified by a unique comment number. Appendix C presents the concerns and NRC staff responses. Each concern embodies one or more comments on similar or related issues. The associated comment numbers are referenced for each concern. The concerns are organized by topic areas. A three-letter identifier for the topic, followed by a number, is assigned to each concern.

  2. US NRC-Sponsored Research on Stress Corrosion Cracking Susceptibility of Dry Storage Canister Materials in Marine Environments - 13344

    SciTech Connect (OSTI)

    Oberson, Greg; Dunn, Darrell [U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington DC, 20555 (United States)] [U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington DC, 20555 (United States); Mintz, Todd; He, Xihua; Pabalan, Roberto; Miller, Larry [Center for Nuclear Waste Regulatory Analyses, 6220 Culebra Rd, San Antonio TX, 78238 (United States)] [Center for Nuclear Waste Regulatory Analyses, 6220 Culebra Rd, San Antonio TX, 78238 (United States)

    2013-07-01T23:59:59.000Z

    At a number of locations in the U.S., spent nuclear fuel (SNF) is maintained at independent spent fuel storage installations (ISFSIs). These ISFSIs, which include operating and decommissioned reactor sites, Department of Energy facilities in Idaho, and others, are licensed by the U.S. Nuclear Regulatory Commission (NRC) under Title 10 of the Code of Federal Regulations, Part 72. The SNF is stored in dry cask storage systems, which most commonly consist of a welded austenitic stainless steel canister within a larger concrete vault or overpack vented to the external atmosphere to allow airflow for cooling. Some ISFSIs are located in marine environments where there may be high concentrations of airborne chloride salts. If salts were to deposit on the canisters via the external vents, a chloride-rich brine could form by deliquescence. Austenitic stainless steels are susceptible to chloride-induced stress corrosion cracking (SCC), particularly in the presence of residual tensile stresses from welding or other fabrication processes. SCC could allow helium to leak out of a canister if the wall is breached or otherwise compromise its structural integrity. There is currently limited understanding of the conditions that will affect the SCC susceptibility of austenitic stainless steel exposed to marine salts. NRC previously conducted a scoping study of this phenomenon, reported in NUREG/CR-7030 in 2010. Given apparent conservatisms and limitations in this study, NRC has sponsored a follow-on research program to more systematically investigate various factors that may affect SCC including temperature, humidity, salt concentration, and stress level. The activities within this research program include: (1) measurement of relative humidity (RH) for deliquescence of sea salt, (2) SCC testing within the range of natural absolute humidity, (3) SCC testing at elevated temperatures, (4) SCC testing at high humidity conditions, and (5) SCC testing with various applied stresses. Results to date indicate that the deliquescence RH for sea salt is close to that of MgCl{sub 2} pure salt. SCC is observed between 35 and 80 deg. C when the ambient (RH) is close to or higher than this level, even for a low surface salt concentration. (authors)

  3. 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 ...

  4. Effect of Hurricane Andrew on the Turkey Point Nuclear Generating Station from August 20--30, 1992. [Final report

    SciTech Connect (OSTI)

    Hebdon, F.J. [Institute of Nuclear Power Operations, Atlanta, GA (United States)

    1993-03-01T23:59:59.000Z

    On August 24, 1992, Hurricane Andrew, a Category 4 hurricane, struck the Turkey Point Electrical Generating Station with sustained winds of 145 mph (233 km/h). This is the report of the team that the US Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO) jointly sponsored (1) to review the damage that the hurricane caused the nuclear units and the utility`s actions to prepare for the storm and recover from it, and (2) to compile lessons that might benefit other nuclear reactor facilities.

  5. PRA In Design: Increasing Confidence in Pre-operational Assessments of Risks (Results of a Joint NASA/ NRC Workshop)

    SciTech Connect (OSTI)

    Robert Youngblood

    2010-06-01T23:59:59.000Z

    In late 2009, the National Aeronautics and Space Administration (NASA) and the U.S. Nuclear Regulatory Commission (NRC) jointly organized a workshop to discuss technical issues associated with application of risk assessments to early phases of system design. The workshop, which was coordinated by the Idaho National Laboratory, involved invited presentations from a number of PRA experts in the aerospace and nuclear fields and subsequent discussion to address the following questions: (a) What technical issues limit decision-makers’ confidence in PRA results, especially at a preoperational phase of the system life cycle? (b) What is being done to address these issues? (c) What more can be done? The workshop resulted in participant observations and suggestions on several technical issues, including the pursuit of non-traditional approaches to risk assessment and the verification and validation of risk models. The workshop participants also identified several important non-technical issues, including risk communication with decision makers, and the integration of PRA into the overall design process.

  6. OVERVIEW OF THE U.S. DEPARTMENT OF ENERGY AND NUCLEAR REGULATORY COMMISSION PERFORMANCE ASSESSMENT APPROACHES: CEMENTITIOUS BARRIERS PARTNERSHIP

    SciTech Connect (OSTI)

    Langton, C.; Burns, H.

    2009-05-29T23:59:59.000Z

    Engineered barriers including cementitious barriers are used at sites disposing or contaminated with low-level radioactive waste to enhance performance of the natural environment with respect to controlling the potential spread of contaminants. Drivers for using cementitious barriers include: high radionuclide inventory, radionuclide characteristics (e.g., long half-live, high mobility due to chemical form/speciation, waste matrix properties, shallow water table, and humid climate that provides water for leaching the waste). This document comprises the first in a series of reports being prepared for the Cementitious Barriers Partnership. The document is divided into two parts which provide a summary of: (1) existing experience in the assessment of performance of cementitious materials used for radioactive waste management and disposal and (2) sensitivity and uncertainty analysis approaches that have been applied for assessments. Each chapter is organized into five parts: Introduction, Regulatory Considerations, Specific Examples, Summary of Modeling Approaches and Conclusions and Needs. The objective of the report is to provide perspective on the state of the practice for conducting assessments for facilities involving cementitious barriers and to identify opportunities for improvements to the existing approaches. Examples are provided in two contexts: (1) performance assessments conducted for waste disposal facilities and (2) performance assessment-like analyses (e.g., risk assessments) conducted under other regulatory regimes. The introductory sections of each section provide a perspective on the purpose of performance assessments and different roles of cementitious materials for radioactive waste management. Significant experience with assessments of cementitious materials associated with radioactive waste disposal concepts exists in the US Department of Energy Complex and the commercial nuclear sector. Recently, the desire to close legacy facilities has created a need to assess the behavior of cementitious materials for applications in environmental remediation and decontamination and decommissioning (D&D) applications. The ability to assess the use and benefits of cementitious materials for these applications can significantly affect decisions related to cleanup activities. For example the need for costly remedial actions may not be necessary if existing or new cementitious barriers were adequately represented. The sections dealing with regulatory considerations include summaries of the different regulations that are relevant for various applications involving cementitious materials. A summary of regulatory guidance and/or policies pertaining to performance assessment of cementitious materials and sensitivity and uncertainty analyses is also provided in the following chapters. Numerous examples of specific applications are provided in each report. The examples are organized into traditional waste disposal applications (performance assessments), applications related to environmental remediation and D&D, and reactor and spent fuel related assessments. Sections that discuss specific facilities or sites contain: (1) descriptions of the role of the cementitious barriers or sensitivity/uncertainty analysis, (2) parameter assumptions and conceptual models, and (3) a relative discussion of the significance in the context of the assessment. Examples from both the U.S. Department of Energy Sites and the U.S. Nuclear Regulatory Commission are provided to illustrate the variety of applications and approaches that have been used. In many cases, minimal credit was taken for cementitious barriers. However, in some of those cases, benefits of being able to take credit for barriers were identified. The examples included: (1) disposal facilities (vaults, trenches, tank closures, cementitious waste forms and containers, etc.), (2) environmental remediation (old disposal facilities), (3) reactor and large structure decommissioning, and (4) spent fuel pools. These examples were selected to provide a perspective on the various ne

  7. Final Environmental Impact Statement for the construction and operation of Claiborne Enrichment Center, Homer, Louisiana (Docket No. 70-3-70). Volume 2, Public comments and NRC response

    SciTech Connect (OSTI)

    Zeitoun, A. [Science Applications International Corp., Germantown, MD (United States)

    1994-08-01T23:59:59.000Z

    The Final Environmental Impact Statement (FEIS) (Volume 1), was prepared by the Nuclear Regulatory Commission (NRC) in accordance with regulation 10 CFR Part 51, which implements the National Environmental Policy Act (NEPA), to assess the potential environmental impacts for licensing the construction and operation of a proposed gaseous centrifuge enrichment facility to be built in Claiborne Parish, Louisiana by Louisiana Energy Services, L.P. (LES). The proposed facility would have a production capacity of about 866 metric tons annually of up to 5 weight percent enriched UF{sub 6}, using a proven centrifuge technology. Included in the assessment are co on, both normal operations and potential accidents (internal and external events), and the eventual decontamination and decommissioning of the site. In order to help assure that releases from the operation of the facility and potential impacts on the public are as low as reasonably achievable, an environmental monitoring program was developed by LES to detect significant changes in the background levels of uranium around the site. Other issues addressed include the purpose and need for the facility, the alternatives to the proposed action, potential disposition of the tails, the site selection process, and environmental justice. The NRC staff concludes that the facility can be constructed and operated with small and acceptable impacts on the public and the environment, and proposes to issue a license to the applicant, Louisiana Energy Services, to authorize construction and operation of the proposed facility. The letters in this Appendix have been divided into three sections. Section One contains letters to which the NRC responded by addressing specific comments. Section Two contains the letters that concerned the communities of Forest Grove and Center Springs. Section Three is composed of letters that required no response. These letters were generally in support of the facility.

  8. NRC comprehensive records disposition schedule. Revision 3

    SciTech Connect (OSTI)

    NONE

    1998-02-01T23:59:59.000Z

    Title 44 US Code, ``Public Printing and Documents,`` regulations issued by the General Service Administration (GSA) in 41 CFR Chapter 101, Subchapter B, ``Management and Use of Information and Records,`` and regulations issued by the National Archives and Records Administration (NARA) in 36 CFR Chapter 12, Subchapter B, ``Records Management,`` require each agency to prepare and issue a comprehensive records disposition schedule that contains the NARA approved records disposition schedules for records unique to the agency and contains the NARA`s General Records Schedules for records common to several or all agencies. The approved records disposition schedules specify the appropriate duration of retention and the final disposition for records created or maintained by the NRC. NUREG-0910, Rev. 3, contains ``NRC`s Comprehensive Records Disposition Schedule,`` and the original authorized approved citation numbers issued by NARA. Rev. 3 incorporates NARA approved changes and additions to the NRC schedules that have been implemented since the last revision dated March, 1992, reflects recent organizational changes implemented at the NRC, and includes the latest version of NARA`s General Records Schedule (dated August 1995).

  9. Guidelines for inservice testing at nuclear power plants

    SciTech Connect (OSTI)

    Campbell, P.

    1995-04-01T23:59:59.000Z

    The staff of the U.S. Nuclear Regulatory Commission (NRC) gives licensees guidelines and recommendations for developing and implementing programs for the inservice testing of pumps and valves at commercial nuclear power plants. The staff discusses the regulations; the components to be included in an inservice testing program; and the preparation and content of cold shutdown justifications, refueling outage justifications, and requests for relief from the American Society of Mechanical Engineers Code requirements. The staff also gives specific guidance on relief acceptable to the NRC and advises licensees in the use of this information at their facilities. The staff discusses the revised standard technical specifications for the inservice testing program requirements and gives guidance on the process a licensee may follow upon finding an instance of noncompliance with the Code.

  10. Measurement of neutron spectra for determining dose equivalent rates at the Texas A&M University Nuclear Science Center

    E-Print Network [OSTI]

    Sanza, Bruce Jerome

    1986-01-01T23:59:59.000Z

    Helium-3 Proportional Counter LET Proportional Counter. METHOD. RESULTS AND DISCUSSION 14 . 15 17 21 25 Multispheres Helium-3 Proportional Counter LET Proportional Counter 25 . . . 30 32 VI CONCLUSIONS 39 REFERENCES. . 41 SUPPLEMENTAL... for performance of personnel neutron dosimeters is provided by the American National Standards Institute (ANSI76) and supplemented by the U. S. Nuclear Regulatory Commision (NRC77). These requirements are outlined in Table 1. The dosimeters available for use...

  11. The Creation of a French Basic Nuclear Installation - Description of the Regulatory Process - 13293

    SciTech Connect (OSTI)

    Mahe, Carole [CEA Marcoule - BP17171 - 30207 Bagnols-Sur-Ceze (France)] [CEA Marcoule - BP17171 - 30207 Bagnols-Sur-Ceze (France); Leroy, Christine [CEA Cadarache - 13108 Saint Paul-Lez-Durance (France)] [CEA Cadarache - 13108 Saint Paul-Lez-Durance (France)

    2013-07-01T23:59:59.000Z

    CEA is a French government-funded technological research organization. It has to build a medium-level waste interim storage facility because the geological repository will not be available until 2025. This interim storage facility, called DIADEM, has to be available in 2017. These wastes are coming from the research facilities for spent fuel reprocessing and the dismantling of the most radioactive parts of nuclear facilities. The CEA handles the waste management by inventorying the needs and updating them regularly. The conception of the facility is mainly based on this inventory. It provides quantity and characteristics of wastes and it gives the production schedule until 2035. Beyond mass and volume, main characteristics of these radioactive wastes are chemical nature, radioisotopes, radioactivity, radiation dose, the heat emitted, corrosive or explosive gas production, etc. These characteristics provide information to study the repository safety. DIADEM mainly consists of a concrete cell, isolated from the outside, wherein stainless steel welded containers are stored, stacked in a vertical position in the racks. DIADEM is scheduled to store three types of 8 mm-thick, stainless steel cylindrical containers with an outside diameter 498 mm and height from 620 to 2120 mm. DIADEM will be a basic nuclear installation (INB in French) because of overall activity of radioactive substances stored. The creation of a French basic nuclear installation is subject to authorization according to the French law No. 2006-686 of 13 June 2006 on Transparency and Security in the Nuclear Field. The authorization takes into account the technical and financial capacities of the licensee which must allow him to conduct his project in compliance with these interests, especially to cover the costs of decommissioning the installation and conduct remediation work, and to monitor and maintain its location site or, for radioactive waste disposal installations, to cover the definitive shut-down, maintenance and surveillance expenditure. The authorization is issued by a decree adopted upon advice of the French Nuclear Safety Authority and after a public enquiry. In accordance with Decree No. 2007-1557 of November 2, 2007, the application is filed with the ministries responsible for nuclear safety and the Nuclear Safety Authority. It consists of twelve files and four records information. The favorable opinion of the Nuclear Safety Authority on the folder is required to start the public inquiry. Once the public inquiry is completed, the building permit is issued by the prefect. (authors)

  12. Microsoft PowerPoint - 5_Pat Smith_NMMSS_2013_Presentation_NRC...

    National Nuclear Security Administration (NNSA)

    Reconciliation Pat Smith PSI NRC Lead NRC Reconciliation NRC Reconciliation requirements per NUREGBR-0007, Rev. 6 All RIS's must submit an MSR (Material Summary Report)...

  13. 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.

  14. Risk Informed Assessment of Regulatory and Design Requirements for Future Nuclear Power Plants - Final Technical Report

    SciTech Connect (OSTI)

    Ritterbusch, Stanley; Golay, Michael; Duran, Felicia; Galyean, William; Gupta, Abhinav; Dimitrijevic, Vesna; Malsch, Marty

    2003-01-29T23:59:59.000Z

    OAK B188 Summary of methods proposed for risk informing the design and regulation of future nuclear power plants. All elements of the historical design and regulation process are preserved, but the methods proposed for new plants use probabilistic risk assessment methods as the primary decision making tool.

  15. The NUCLARR databank: Human reliability and hardware failure data for the nuclear power industry

    SciTech Connect (OSTI)

    Reece, W.J.

    1993-05-01T23:59:59.000Z

    Under the sponsorship of the US Nuclear Regulatory Commission (NRC), the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) was developed to provide human reliability and hardware failure data to analysts in the nuclear power industry. This IBM-compatible databank is contained on a set of floppy diskettes which include data files and a menu-driven system for locating, reviewing, sorting, and retrieving the data. NUCLARR contains over 2500 individual data records, drawn from more, than 60 sources. The system is upgraded annually, to include additional human error and hardware component failure data and programming enhancements (i.e., increased user-friendliness). NUCLARR is available from the NRC through project staff at the INEL.

  16. Technical Support for Improving the Licensing Regulatory Base for Selected Facilities Associated with the Front End of the Fuel Cycle

    SciTech Connect (OSTI)

    Clark, R. G.; Schreiber, R. E.; Jamison, J. D.; Davenport, L. C.; Brite, D. W.

    1982-04-01T23:59:59.000Z

    Pacific Northwest Laboratory (PNL) was asked by the NRC Office of Nuclear Material Safety and Safeguards (NMSS) to determine the adequacy of its health, safety and environmental regulatory base as a guide to applicants for licenses to operate UF{sub 6} conversion facilities and fuel fabrication plants. The regulatory base was defined as the body of documented requirements and guidance to licensees, including laws passed by Congress, Federal Regulations developed by the NRC to implement the laws, license conditions added to each license to deal with special requirements for that specific license, and Regulatory Guides. The study concentrated on the renewal licensing accomplished in the last few years at five typical facilities, and included analyses of licensing documents and interviews with individuals involved with different aspects of the licensing process. Those interviewed included NMSS staff, Inspection and Enforcement (IE) officials, and selected licensees. From the results of the analyses and interviews, the PNL study team concludes that the regulatory base is adequate but should be codified for greater visibility. PNL recommends that NMSS clarify distinctions among legal requirements of the licensee, acceptance criteria employed by NMSS, and guidance used by all. In particular, a prelicensing conference among NMSS, IE and each licensee would be a practical means of setting license conditions acceptable to all parties.

  17. Report to the Nuclear Regulatory Commission from the staff panel on the Commission's determination of an Extraordinary Nuclear Occurrence (ENO)

    SciTech Connect (OSTI)

    none,

    1980-01-01T23:59:59.000Z

    The Panel finds that the first criterion, pertaining to whether the accident caused a discharge of radioactive material or levels of radiation offsite as defined in 10 CFR 140.84, has not been met. It further finds that there is presently insufficient information to support any definitive finding as to whether or not the second criterion, relating to damage to persons or property offsite as defined in 10 CFR 140.85, has been met. Since the Panel has not found that both criteria have been met, it recommends that the Commission determine that the accident at Three Mile Island did not constitute an extraordinary nuclear occurrence.

  18. Two dose-estimation models CSA-N288.1 and Nureg 1.109, 1.113 - compared for chronic aquatic releases from nuclear facilities

    E-Print Network [OSTI]

    Sheppard, S C; Peterson, S R

    2000-01-01T23:59:59.000Z

    Both the Canadian Standards Association (CSA) and the United States Nuclear Regulatory Commission (US-NRC) have published guidelines for the calculation of doses to the public due to emissions from nuclear facilities. In the sale of CANDU reactors overseas, either of these guidelines may be used as part of the approval process in the recipient country. This study compares the aquatic exposure pathways described in the guidelines. These include direct consumption of contaminated water and food, and exposure to contaminated sediments. The CSA and US-NRC guidelines for estimating dilution of aquatic emissions are of a general nature and the choice of model used to quantify dilution is left to the user. The models prescribed for the different exposure pathways by these two regulatory guides are similar in many attributes. Many of the recommended parameter values are identical and many of the formulations are either identical, or become identical under general conditions. However, despite these similarities, there...

  19. Site-specific EIS ordered but injunctive relief deined in nuclear waste storage case

    SciTech Connect (OSTI)

    Barnhart y Chavez, S.

    1980-01-01T23:59:59.000Z

    The Energy Research and Development Administration (ERDA) received appropriations in 1976-77 to construct 22 tanks for storage of high level radioactive wastes generated by its nuclear weapons materials production program. The tanks were to replace older, leaking tanks at the Hanford Reservation in Richland, Washington and the Savannah River Plant in Aiken, South Carolina. The Natural Resources Defense Council (NRDC) had unsuccessfully requested that ERDA obtain a construction permit from the Nuclear Regulatory Commission (NRC). NRDC also petitioned NRC to exercise its licensing authority over the tanks under Section 202(4) of the Energy Reorganization Act of 1974. In response to the NRDC request, ERDA claimed the tanks were only for short-term storage and therefore a license was unnecessary. NRC claimed it lacked jurisdiction over the tanks. NRDC filed suit in United States District Court for the District of Columbia, alleging that ERDA had violated Section 102(2)(C) of the National Environmental Policy Act, and that both ERDA and NRC had violated Section 202(4) of the Energy Reorganization Act. NRDC requested an injunction against further construction of the tanks. Although ERDA did not have to obtain an NRC construction permit for the nuclear waste storage tanks at Hanford Reservation and Savannah River Plant, the programmatic Environmental Impact Statement submitted was insufficient and site-specific statements must be prepared. Injunctive relief pending the statements was denied for the social and economic costs of delaying the tanks project. NRC decisions even remotely connected to its licensing power should be contested in federal courts of appeals, not district courts. The court gave NRDC a hollow victory by ordering a more specific EIS, but denying an injunction.

  20. Qualification of NDE personnel in the nuclear industry

    SciTech Connect (OSTI)

    Epps, T.N.

    1984-06-01T23:59:59.000Z

    There has been evidence of ineffective programs for certifying nondestructive examination (NDE) personnel who conduct periodic inservice examinations in nuclear power plants under ASME Section XI Code requirements. This was brought to the attention of a group from the electric utility industry, the Electric Power Research Institute (EPRI), some NDE consultants and representatives from the American Society of Mechanical Engineers (ASME) by the Nuclear Regulatory Commission (NRC) in a May, 1982 meeting in Bethesda, Maryland. One problem pointed out by the NRC was the lack of a clear definition of qualification requirements for certification of NDE personnel who conduct ASME Section XI Inservice Inspection work in nuclear power plants. The NRC requested that the nuclear industry resolve this problem by formulating definitive qualification requirements for personnel certification that could be made an industry requirement. In June, 1982 the EPRI NDE Subcommittee held a general meeting for utility representatives to discuss the results of the May, 1982 meeting to develop a plan for industry response to the issue. The consensus was that an Ad Hoc Committee of utility representatives be convened to develop a document outlining qualification requirements for vertification of NDE personnel. The Ad Hoc Committee was formally convened on September 29, 1982.

  1. Digital upgrade issues and the evolving regulatory environment

    SciTech Connect (OSTI)

    Meininger, R. D. [CHAR Services, Inc., 400 E. Main St., Annville, PA 17003 (United States)

    2006-07-01T23:59:59.000Z

    This paper deals with the qualification of an Instrumentation and Control (I and C) upgrade for Electromagnetic Compatibility (EMC) in the plant, focusing on the interpretation of the NRC Regulatory Guide 1.180 Revision 1, 'Guidelines for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related Instrumentation and Control Systems.' Options presented by Reg. Guide 1.180 are discussed along with alternative EMC Guidelines being used by nuclear power plants. Problems commonly encountered during the EMC qualification process are discussed and suggestions presented on how to deal with these common problems. Also included is a discussion of an emerging issue of how to address the issue of EMC of replacement discrete modules or printer circuit (PC) boards in a system that was either previously qualified or never qualified for EMC. (authors)

  2. Survey of Field Programmable Gate Array Design Guides and Experience Relevant to Nuclear Power Plant Applications

    SciTech Connect (OSTI)

    Bobrek, Miljko [ORNL; Bouldin, Don [ORNL; Holcomb, David Eugene [ORNL; Killough, Stephen M [ORNL; Smith, Stephen Fulton [ORNL; Ward, Christina D [ORNL

    2007-09-01T23:59:59.000Z

    From a safety perspective, it is difficult to assess the correctness of FPGA devices without extensive documentation, tools, and review procedures. NUREG/CR-6463, "Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems," provides guidance to the Nuclear Regulatory Commission (NRC) on auditing of programs for safety systems written in ten high-level languages. A uniform framework for the formulation and discussion of language-specific programming guidelines was employed. Comparable guidelines based on a similar framework are needed for FPGA-based systems. The first task involves evaluation of regulatory experience gained by other countries and other agencies, and those captured in existing standards, to identify regulatory approaches that can be adopted by NRC. If existing regulations do not provide a sufficient regulatory basis for adopting relevant regulatory approaches that are uncovered, ORNL will identify the gaps. Information for this report was obtained through publicly available sources such as published papers and presentations. No proprietary information is represented.

  3. 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.

  4. 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.

  5. Shutdown and low-power operation at commercial nuclear power plants in the United States. Final report

    SciTech Connect (OSTI)

    Not Available

    1993-09-01T23:59:59.000Z

    The report contains the results of the NRC Staff`s evaluation of shutdown and low-power operations at US commercial nuclear power plants. The report describes studies conducted by the staff in the following areas: Operating experience related to shutdown and low-power operations, probabilistic risk assessment of shutdown and low-power conditions and utility programs for planning and conducting activities during periods the plant is shut down. The report also documents evaluations of a number of technical issues regarding shutdown and low-power operations performed by the staff, including the principal findings and conclusions. Potential new regulatory requirements are discussed, as well as potential changes in NRC programs. A draft report was issued for comment in February 1992. This report is the final version and includes the responses to the comments along with the staff regulatory analysis of potential new requirements.

  6. NRC safety research in support of regulation. Volume 8, FY 1993

    SciTech Connect (OSTI)

    Not Available

    1994-06-01T23:59:59.000Z

    This report, the ninth in a series of annual reports, was prepared in response to congressional inquiries concerning how nuclear regulatory research is used. It summarizes the accomplishments of the Office of Nuclear Regulatory Research during FY 1993. A special emphasis on accomplishments in nuclear power plant aging research reflects recognition that number of plants are entering the final portion of their original 40-year operating licenses and that, in addition to current aging effects, a focus on safety considerations for license renewal becomes timely. The primary purpose of performing regulatory research is to develop and provide the Commission and its staff with sound technical bases for regulatory decisions on the safe operation of licensed nuclear reactors and facilities, to find unknown or unexpected safety problems, and to develop data and related information for the purpose of revising the Commission`s rules, regulatory guides, or other guidance.

  7. o:\informs\fixforms\nrc741.wpf

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnual Siteof Energy 2, 2015Visiting8.pdfand Characterization of aHome *NRC FORM 741 (5-2002)

  8. Coping with nuclear power risks: the electric utility incentives

    SciTech Connect (OSTI)

    Starr, C.; Whipple, C.

    1982-01-01T23:59:59.000Z

    The financial risks associated with nuclear power accidents are estimated by interpolating between frequency-vs.-severity data from routine outages and the frequency-vs.-severity estimates from the Nuclear Regulatory Commission's (NRC's) Reactor Safety Study (WASH-1400). This analysis indicates that the expected costs of plant damage and lost power production are large compared to the public risks estimated in WASH-1400, using values from An Approach to Quantitative Safety Goals for Nuclear Power Plants (NUREG-0739), prepared by the NRC Advisory Committee on Reactor Safeguards. Analyses of the cost-effectiveness of accident-prevention investments that include only anticipated public safety benefits will underestimate the value of such investments if reductions in power plant damage risk are not included. The analysis also suggests that utility self-interest and the public interest in safety are generally coincident. It is argued that greater use could be made of this self-interest in regulation if the relationship between the NRC and the industry were more cooperative, less adversary in nature.

  9. Independent Verification and Validation Of SAPHIRE 8 Software Requirements Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Kent Norris

    2010-03-01T23:59:59.000Z

    The purpose of the Independent Verification and Validation (IV&V) role in the evaluation of the SAPHIRE requirements definition is to assess the activities that results in the specification, documentation, and review of the requirements that the software product must satisfy, including functionality, performance, design constraints, attributes and external interfaces. The IV&V team began this endeavor after the software engineering and software development of SAPHIRE had already been in production. IV&V reviewed the requirements specified in the NRC Form 189s to verify these requirements were included in SAPHIRE’s Software Verification and Validation Plan (SVVP).

  10. Independent Verification and Validation Of SAPHIRE 8 Software Requirements Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Kent Norris

    2009-09-01T23:59:59.000Z

    The purpose of the Independent Verification and Validation (IV&V) role in the evaluation of the SAPHIRE requirements definition is to assess the activities that results in the specification, documentation, and review of the requirements that the software product must satisfy, including functionality, performance, design constraints, attributes and external interfaces. The IV&V team began this endeavor after the software engineering and software development of SAPHIRE had already been in production. IV&V reviewed the requirements specified in the NRC Form 189s to verify these requirements were included in SAPHIRE’s Software Verification and Validation Plan (SVVP).

  11. Development of Risk Insights for Regulatory Review of a Near-Surface Disposal Facility for Radioactive Waste

    SciTech Connect (OSTI)

    Esh, D.W.; Ridge, A.C.; Thaggard, M. [U.S. Nuclear Regulatory Commission, Mail Stop T7J8, Washington, DC 20555 (United States)

    2006-07-01T23:59:59.000Z

    Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the Department of Energy (DOE) to consult with the Nuclear Regulatory Commission (NRC) about non-High Level Waste (HLW) determinations. In its consultative role, NRC performs technical reviews of DOE's waste determinations but does not have regulatory authority over DOE's waste disposal activities. The safety of disposal is evaluated by comparing predicted disposal facility performance to the performance objectives specified in NRC regulations for the disposal of low-level waste (10 CFR Part 61 Subpart C). The performance objectives contain criteria for protection of the public, protection of inadvertent intruders, protection of workers, and stability of the disposal site after closure. The potential radiological dose to receptors typically is evaluated with a performance assessment (PA) model that simulates the release of radionuclides from the disposal site, transport of radionuclides through the environment, and exposure of potential receptors to residual contamination for thousands of years. This paper describes NRC's development and use of independent performance assessment modeling to facilitate review of DOE's non-HLW determination for the Saltstone Disposal Facility (SDF) at the Savannah River Site. NRC's review of the safety of near-surface disposal of radioactive waste at the SDF was facilitated and focused by risk insights developed with an independent PA model. The main components of NRC's performance assessment model are presented. The development of risk insights that allow the staff to focus review efforts on those areas that are most important to satisfying the performance objectives is discussed. Uncertainty analysis was performed of the full stochastic model using genetic variable selection algorithms. The results of the uncertainty analysis were then used to guide the development of simulations of other scenarios to understand the key risk drivers and risk limiters of the SDF. Review emphasis was placed on those aspects of the disposal system that were expected to drive performance: the physical and chemical performance of the cementitious wasteform and concrete vaults. Refinement of the modeling of the degradation and release from the cementitious wasteform had a significant effect on the predicted dose to a member of the public. (authors)

  12. A RE-LOOK AT THE US NRC SAFETY GOALS

    SciTech Connect (OSTI)

    mubayi v.

    2013-09-22T23:59:59.000Z

    Since they were adopted in 1986, the US NRC’s Safety Goals have played a valuable role as a de facto risk acceptance criterion against which the predicted performance of a commercial nuclear power reactor can be evaluated and assessed. The current safety goals are cast in terms of risk metrics called quantitative health objectives (QHOs), limiting numerical values of the risks of the early and latent health effects of accidental releases of radioactivity to the offsite population. However, while demonstrating compliance with current safety goals has been an important step in assessing the acceptance of the risk posed by LWRs, new or somewhat different goals may be needed that go beyond the current early fatality and latent cancer fatality QHOs in assessing reactor risk. Natural phenomena such as hurricanes seem to be suitable candidates for establishing a background rate to derive a risk goal as their order of magnitude cost of damages is similar to those estimated in severe accident Level 3 PRAs done for nuclear power plants. This paper obtains a risk goal that could have a wider applicability, compared to the current QHOs, as a technology-neutral goal applicable to future reactors and multi-unit sites.

  13. Nuclear Regulatory Commission issuances

    SciTech Connect (OSTI)

    NONE

    1997-09-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM). The summaries and headnotes preceding the opinions reported herein are not to be deemed a part of those opinions or have any independent legal significance.

  14. Nuclear Regulatory Commission Issuances

    SciTech Connect (OSTI)

    NONE

    1996-12-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM).

  15. NUCLEAR REGULATORY COMMISSION

    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 Strategic2 OPAM615_Cost Estimating35.docMusings on| Department of0Register: December 21,

  16. Part 1, Use of seismic experience and test data to show ruggedness of equipment in nuclear power plants; Part 2, Review procedure to assess seismic ruggedness of cantilever bracket cable tray supports

    SciTech Connect (OSTI)

    Kennedy, R.P. (Structural Mechanics Consulting, Inc., Yorba Linda, CA (United States)); von Riesemann, W.A. (Sandia National Labs., Albuquerque, NM (United States)); Wyllie, L.A. Jr. (Degenkolb (H.J.) Associates, San Francisco, CA (United States)); Schiff, A.J. (Stanford Univ., CA (United States)); Ibanez, P. (Anco Engineers, Inc., Culver City, CA (United States))

    1992-06-01T23:59:59.000Z

    In December 1980, the US Nuclear Regulatory Commission (NRC) designated Seismic Qualification of Equipment in Operating Plants'' as an Unresolved Safety Issue (USI), A-46. The objective of USI A-46 is to develop alternative seismic qualification methods and acceptance criteria that can be used to assess the capability of mechanical and electrical equipment in operating nuclear power plants to perform the intended safety functions. A group of affected utilities formed the Seismic Qualification Utility Group (SQUG) to work with the NRC in developing a program methodology to enable resolution of the A-46 issue. To assist in developing a program methodology, SQUG and the NRC jointly selected and supported a five-member Senior Seismic Review and Advisory Panel (SSRAP) in June 1983 to make an independent assessment of whether certain classes of equipment in operating nuclear power plants in the United States have demonstrated sufficient ruggedness in past earthquakes so as to render an explicit seismic qualification unnecessary. SSRAP operated as an independent review body with all of its findings submitted concurrently to both SQUG and the NRC. During their period of involvement, SSRAP issued several draft reports on their conclusions. This document contains the final versions of these reports; namely, Use of Seismic Experience and Test Data to Show Ruggedness of Equipment in Nuclear Power Plants,'' dated February 1991 and Review Procedure to Assess Seismic Ruggedness of Cantilever Bracket Cable Tray Supports,'' dated March 1, 1991.

  17. Part 1, Use of seismic experience and test data to show ruggedness of equipment in nuclear power plants; Part 2, Review procedure to assess seismic ruggedness of cantilever bracket cable tray supports

    SciTech Connect (OSTI)

    Kennedy, R.P. [Structural Mechanics Consulting, Inc., Yorba Linda, CA (United States); von Riesemann, W.A. [Sandia National Labs., Albuquerque, NM (United States); Wyllie, L.A. Jr. [Degenkolb (H.J.) Associates, San Francisco, CA (United States); Schiff, A.J. [Stanford Univ., CA (United States); Ibanez, P. [Anco Engineers, Inc., Culver City, CA (United States)

    1992-06-01T23:59:59.000Z

    In December 1980, the US Nuclear Regulatory Commission (NRC) designated ``Seismic Qualification of Equipment in Operating Plants`` as an Unresolved Safety Issue (USI), A-46. The objective of USI A-46 is to develop alternative seismic qualification methods and acceptance criteria that can be used to assess the capability of mechanical and electrical equipment in operating nuclear power plants to perform the intended safety functions. A group of affected utilities formed the Seismic Qualification Utility Group (SQUG) to work with the NRC in developing a program methodology to enable resolution of the A-46 issue. To assist in developing a program methodology, SQUG and the NRC jointly selected and supported a five-member Senior Seismic Review and Advisory Panel (SSRAP) in June 1983 to make an independent assessment of whether certain classes of equipment in operating nuclear power plants in the United States have demonstrated sufficient ruggedness in past earthquakes so as to render an explicit seismic qualification unnecessary. SSRAP operated as an independent review body with all of its findings submitted concurrently to both SQUG and the NRC. During their period of involvement, SSRAP issued several draft reports on their conclusions. This document contains the final versions of these reports; namely, ``Use of Seismic Experience and Test Data to Show Ruggedness of Equipment in Nuclear Power Plants,`` dated February 1991 and ``Review Procedure to Assess Seismic Ruggedness of Cantilever Bracket Cable Tray Supports,`` dated March 1, 1991.

  18. The Environmental Protection Agency's Safety Standards for Disposal of Spent Nuclear Fuel: Potential Path Forward in Response to the Report of the Blue Ribbon Commission on America's Nuclear Future - 13388

    SciTech Connect (OSTI)

    Forinash, Betsy; Schultheisz, Daniel; Peake, Tom [U.S. Environmental Protection Agency, Radiation Protection Division (United States)] [U.S. Environmental Protection Agency, Radiation Protection Division (United States)

    2013-07-01T23:59:59.000Z

    Following the decision to withdraw the Yucca Mountain license application, the Department of Energy created a Blue Ribbon Commission (BRC) on America's Nuclear Future, tasked with recommending a national strategy to manage the back end of the nuclear fuel cycle. The BRC issued its final report in January 2012, with recommendations covering transportation, storage and disposal of spent nuclear fuel (SNF); potential reprocessing; and supporting institutional measures. The BRC recommendations on disposal of SNF and high-level waste (HLW) are relevant to the U.S. Environmental Protection Agency (EPA), which shares regulatory responsibility with the Nuclear Regulatory Commission (NRC): EPA issues 'generally applicable' performance standards for disposal repositories, which are then implemented in licensing. For disposal, the BRC endorses developing one or more geological repositories, with siting based on an approach that is adaptive, staged and consent-based. The BRC recommends that EPA and NRC work cooperatively to issue generic disposal standards-applying equally to all sites-early in any siting process. EPA previously issued generic disposal standards that apply to all sites other than Yucca Mountain. However, the BRC concluded that the existing regulations should be revisited and revised. The BRC proposes a number of general principles to guide the development of future regulations. EPA continues to review the BRC report and to assess the implications for Agency action, including potential regulatory issues and considerations if EPA develops new or revised generic disposal standards. This review also involves preparatory activities to define potential process and public engagement approaches. (authors)

  19. Human Factors Considerations in New Nuclear Power Plants: Detailed Analysis.

    SciTech Connect (OSTI)

    OHara,J.; Higgins, J.; Brown, W.; Fink, R.

    2008-02-14T23:59:59.000Z

    This Nuclear Regulatory Commission (NRC) sponsored study has identified human-performance issues in new and advanced nuclear power plants. To identify the issues, current industry developments and trends were evaluated in the areas of reactor technology, instrumentation and control technology, human-system integration technology, and human factors engineering (HFE) methods and tools. The issues were organized into seven high-level HFE topic areas: Role of Personnel and Automation, Staffing and Training, Normal Operations Management, Disturbance and Emergency Management, Maintenance and Change Management, Plant Design and Construction, and HFE Methods and Tools. The issues where then prioritized into four categories using a 'Phenomena Identification and Ranking Table' methodology based on evaluations provided by 14 independent subject matter experts. The subject matter experts were knowledgeable in a variety of disciplines. Vendors, utilities, research organizations and regulators all participated. Twenty issues were categorized into the top priority category. This Brookhaven National Laboratory (BNL) technical report provides the detailed methodology, issue analysis, and results. A summary of the results of this study can be found in NUREG/CR-6947. The research performed for this project has identified a large number of human-performance issues for new control stations and new nuclear power plant designs. The information gathered in this project can serve as input to the development of a long-term strategy and plan for addressing human performance in these areas through regulatory research. Addressing human-performance issues will provide the technical basis from which regulatory review guidance can be developed to meet these challenges. The availability of this review guidance will help set clear expectations for how the NRC staff will evaluate new designs, reduce regulatory uncertainty, and provide a well-defined path to new nuclear power plant licensing.

  20. Design-Basis Flood Estimation for Site Characterization at Nuclear Power Plants in the United States of America

    SciTech Connect (OSTI)

    Prasad, Rajiv; Hibler, Lyle F.; Coleman, Andre M.; Ward, Duane L.

    2011-11-01T23:59:59.000Z

    The purpose of this document is to describe approaches and methods for estimation of the design-basis flood at nuclear power plant sites. Chapter 1 defines the design-basis flood and lists the U.S. Nuclear Regulatory Commission's (NRC) regulations that require estimation of the design-basis flood. For comparison, the design-basis flood estimation methods used by other Federal agencies are also described. A brief discussion of the recommendations of the International Atomic Energy Agency for estimation of the design-basis floods in its member States is also included.

  1. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  2. NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669

    SciTech Connect (OSTI)

    Not Available

    1994-08-01T23:59:59.000Z

    The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

  3. Independent Verification and Validation Of SAPHIRE 8 Software Quality Assurance Plan Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Kent Norris

    2010-02-01T23:59:59.000Z

    This report provides an evaluation of the Software Quality Assurance Plan. The Software Quality Assurance Plan is intended to ensure all actions necessary for the software life cycle; verification and validation activities; documentation and deliverables; project management; configuration management, nonconformance reporting and corrective action; and quality assessment and improvement have been planned and a systematic pattern of all actions necessary to provide adequate confidence that a software product conforms to established technical requirements; and to meet the contractual commitments prepared by the sponsor; the Nuclear Regulatory Commission.

  4. Independent Verification and Validation Of SAPHIRE 8 Software Quality Assurance Plan Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Kent Norris

    2010-03-01T23:59:59.000Z

    This report provides an evaluation of the Software Quality Assurance Plan. The Software Quality Assurance Plan is intended to ensure all actions necessary for the software life cycle; verification and validation activities; documentation and deliverables; project management; configuration management, nonconformance reporting and corrective action; and quality assessment and improvement have been planned and a systematic pattern of all actions necessary to provide adequate confidence that a software product conforms to established technical requirements; and to meet the contractual commitments prepared by the sponsor; the Nuclear Regulatory Commission.

  5. Nuclear reactors built, being built, or planned 1993

    SciTech Connect (OSTI)

    Not Available

    1993-08-01T23:59:59.000Z

    Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1993. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: civilian, production, military, export, and critical assembly.

  6. NRC Transportation Security (Part 73 SNF Update and Part 37 Category...

    Office of Environmental Management (EM)

    NRC Transportation Security (Part 73 SNF Update and Part 37 Category 1 and 2 Materials) NRC Transportation Security (Part 73 SNF Update and Part 37 Category 1 and 2 Materials) NRC...

  7. Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.

    SciTech Connect (OSTI)

    LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

    2013-05-01T23:59:59.000Z

    The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

  8. Approach of Czech regulatory body to LBB

    SciTech Connect (OSTI)

    Tendera, P.

    1997-04-01T23:59:59.000Z

    At present there are two NPPs equipped with PWR units in Czech Republic. The Dukovany, NPP is about ten years in operation (four units 440 MW - WWBFL model 213) and Tomelin NPP is under construction (two units 1000 MW - WWER model 320). Both NPPs were built to Soviet design and according to Soviet regulations and standards but most of equipment for primary circuits was supplied by home manufacturers. The objective of the Czech LBB program is to prove the LBB status of the primary piping systems of there NPPs and the LBB concept is a part of strategy to meet western style safety standards. The reason for the Czech LBB project is a lack of some standard safety Facilities too. For both Dukovany and Tomelin NPPs a full LBB analysis should be carried out. The application of LBB to the piping system should be also a cost effective means to avoid installations of pipe whip restraints and jet shields. The Czech regulatory body issued non-mandatory requirement, {open_quotes}Leak Before Break{close_quotes} which is in compliance with national legal documents and which is based on the US NRC Regulatory Procedures and US standards (ASMF CODE, ANSI). The requirement has been published in the document {open_quotes}Safety of Nuclear Facilities{close_quotes} No 1/1991 as {open_quotes}Requirements on the Content and Format of Safety Reports and their Supplements{close_quote} and consist of two parts (1) procedure for obtaining proof of evidence {open_quotes}Leak Before Break{close_quotes} (2) leak detection systems for the pressurized reactor primary circuit. At present some changes concerning both parts of the above document will be introduced. The reasons for this modifications will be presented.

  9. Approach for Czech regulatory body to LBB

    SciTech Connect (OSTI)

    Tendera, P. [State Office for Nuclear Safety (SONS), Prague (Czech Republic)

    1997-04-01T23:59:59.000Z

    At present there are two NPPs equipped with PWR units in Czech Republic. The Dukovany NPP is about ten years in operation (four units 440 MW - WWER model 213) and Temelin NPP is under construction (two units 1000 MW-WWER model 320). Both NPPs were built to Soviet design and according to Soviet regulations and standards but most of equipment for primary circuits was supplied by home manufactures. The objective for the Czech LBB programme is to prove the LBB status of the primary piping systems of these NPPs and the LBB concept is a part of strategy to meet western style safety standards. The reason for the Czech LBB project is a lack of some standard safety facilities, too. For both Dukovany and Temolin NPPs a full LBB analysis should be carried out. The application of LBB to the piping system should be also a cost effective means to avoid installations of pipe whip restraints and jet shields. The Czech regulatory body issued non-mandatory requirement {open_quotes}Leak Before Break{close_quotes} which is in compliance with national legal documents and which is based on the US NRC Regulatory Procedures and US standards (ASME, CODE, ANSI). The requirement has been published in the document {open_quotes}Safety of Nuclear Facilities{close_quotes} No. 1/1991 as {open_quotes}Requirements on the Content and Format of Safety Reports and their Supplements{close_quotes} and consists of two parts (1) procedure for obtaining proof of evidence {open_quotes}Leak Before Break{close_quotes} (2) leak detection systems for the pressurized reactor primary circuit. At present some changes concerning both parts of the above document will be introduced. The reasons for this modifications will be presented.

  10. US nuclear power plant operating cost and experience summaries

    SciTech Connect (OSTI)

    Kohn, W.E.; Reid, R.L.; White, V.S.

    1998-02-01T23:59:59.000Z

    NUREG/CR-6577, U.S. Nuclear Power Plant Operating Cost and Experience Summaries, has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants. Cost incurred after initial construction are characterized as annual production costs, representing fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operating summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from annual operating reports submitted by the licensees, plant histories contained in Nuclear Power Experience, trade press articles, and the Nuclear Regulatory Commission (NRC) web site (www.nrc.gov).

  11. Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2008

    SciTech Connect (OSTI)

    U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research

    2009-12-01T23:59:59.000Z

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2008 annual reports submitted by five of the seven categories1 of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Because there are no geologic repositories for high-level waste currently licensed and no low-level waste disposal facilities in operation, only five categories will be considered in this report.

  12. 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.

  13. The MacArthur Maze Fire and Roadway Collapse: A "Worst Case Scenario" for Spent Nuclear Fuel Transportation?

    SciTech Connect (OSTI)

    Bajwa, Christopher S.; Easton, Earl P.; Adkins, Harold E.; Cuta, Judith M.; Klymyshyn, Nicholas A.; Suffield, Sarah R.

    2012-07-06T23:59:59.000Z

    In 2007, a severe transportation accident occurred near Oakland, California, at the interchange known as the "MacArthur Maze." The accident involved a double tanker truck of gasoline overturning and bursting into flames. The subsequent fire reduced the strength of the supporting steel structure of an overhead interstate roadway causing the collapse of portions of that overpass onto the lower roadway in less than 20 minutes. The US Nuclear Regulatory Commission has analyzed what might have happened had a spent nuclear fuel transportation package been involved in this accident, to determine if there are any potential regulatory implications of this accident to the safe transport of spent nuclear fuel in the United States. This paper provides a summary of this effort, presents preliminary results and conclusions, and discusses future work related to the NRC's analysis of the consequences of this type of severe accident.

  14. Initiating Event Rates at U.S. Nuclear Power Plants 1988–2013

    SciTech Connect (OSTI)

    John A. Schroeder; Gordon R. Bower

    2014-02-01T23:59:59.000Z

    Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants since each plant’s low-power license date. The evaluation is based on the operating experience from fiscal year 1988 through 2013 as reported in licensee event reports. Engineers with nuclear power plant experience staff reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), sequence of events starts when reactor is critical and at or above the point of adding heat, occurs at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and is reported by a licensee event report. This report displays occurrence rates (baseline frequencies) for the categories of initiating events that contribute to the NRC’s Industry Trends Program. Sixteen initiating event groupings are trended and displayed. Initiators are plotted separately for initiating events with different occurrence rates for boiling water reactors and pressurized water reactors. p-values are given for the possible presence of a trend over the most recent 10 years.

  15. 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...

  16. Nuclear Material Transaction Report NRC 741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1 1

  17. Nuclear Material Transaction Report NRC 741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1 1D

  18. Nuclear Material Transaction Report NRC741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1

  19. Nuclear Material Transaction Report nrc741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1 A

  20. Nuclear Material Transaction Report nrc741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1 A

  1. Nuclear Material Transaction Report nrc741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1 A

  2. Nuclear Material Transaction Report nrc741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1 A

  3. Nuclear Material Transaction Report nrc741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1 A

  4. Nuclear Material Transaction Report nrc741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1 A

  5. Nuclear Material Transaction Report nrc741_1

    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 AprilAElectronic Input Options Gary L. Hirsch SNL 2001a,Summary; i-C C l l a10B 1 1 1

  6. DOE/NRC Forms | 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: 5-13-14Russianvolunteer |At.<ENDMENT/MODIFICATION NO.Template FORNRC

  7. Controlling the atom. The beginnings of nuclear regulation 1946--1962

    SciTech Connect (OSTI)

    Mazuzan, G.T.; Walker, J.S.

    1997-08-01T23:59:59.000Z

    This book traces the early history of nuclear power regulation in the US. It focuses on the Atomic Energy Commission (AEC), the federal agency that until 1975 was primarily responsible for planning and carrying out programs to protect public health and safety from the hazards of the civilian use of nuclear energy. It also describes the role of other groups that figured significantly in the development of regulatory policies, including the congressional Joint Committee on Atomic Energy, federal agencies other than the AEC, state governments, the nuclear industry, and scientific organizations. And it considers changes in public perceptions of and attitudes toward atomic energy and the dangers of radiation exposure. The context in which regulatory programs evolved is a rich and complex mixture of political, legislative, legal, technological, scientific, and administrative history. The basic purpose of this book is to provide the Nuclear Regulatory Commission (NRC), which inherited responsibility for nuclear safety after Congress disbanded the AEC, and the general public with information on the historical antecedents and background of regulatory issues.

  8. Energy Department Issues Draft Request For Proposal for Nuclear...

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

    Management Post FY 2015 Acquisition Planning. The DRFP is for managing Spent Nuclear Fuel (SNF) storage facilities and licenses under NRC regulations. The scope...

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

    E-Print Network [OSTI]

    Yen, W.W.S.

    2010-01-01T23:59:59.000Z

    ABBREVIATIONS AAB Accident Analysis Branch (NRC) ApplicationF, Nischan, US NRC. Accident Analysis Branch, U. S. Nuclear2. L 2 is placed in the Accident Analysis Branch (AAB); with

  10. Nuclear reactors built, being built, or planned 1996

    SciTech Connect (OSTI)

    NONE

    1997-08-01T23:59:59.000Z

    This publication contains unclassified information about facilities, built, being built, or planned in the United States for domestic use or export as of December 31, 1996. The Office of Scientific and Technical Information, U.S. Department of Energy, gathers this information annually from Washington headquarters, and field offices of DOE; from the U.S. Nuclear Regulatory Commission (NRC); from the U. S. reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from U.S. and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled.

  11. Dose reduction and optimization studies (ALARA) at nuclear power facilities. [PWR; BWR

    SciTech Connect (OSTI)

    Baum, J.W.; Meinhold, C.B.

    1983-01-01T23:59:59.000Z

    Brookhaven National Laboratory (BNL) has been commissioned by the Nuclear Regulatory Commission (NRC) to study dose-reduction techniques and effectiveness of as low as reasonably achievable (ALARA) planning at LWR plants. These studies have the following objectives: identify high-dose maintenance tasks; identify dose-reduction techniques; examine incentives for dose reduction; evaluate cost-effectiveness and optimization of dose-reduction techniques; and compile an ALARA handbook on data, engineering modifications, cost-effectiveness calculations, and other information of interest to ALARA practioners.

  12. Regulatory Safety Issues in the Structural Design Criteria of ASME Section III Subsection NH and for Very High Temperatures for VHTR & GEN IV

    SciTech Connect (OSTI)

    William J. O’Donnell; Donald S. Griffin

    2007-05-07T23:59:59.000Z

    The objective of this task is to identify issues relevant to ASME Section III, Subsection NH [1], and related Code Cases that must be resolved for licensing purposes for VHTGRs (Very High Temperature Gas Reactor concepts such as those of PBMR, Areva, and GA); and to identify the material models, design criteria, and analysis methods that need to be added to the ASME Code to cover the unresolved safety issues. Subsection NH was originally developed to provide structural design criteria and limits for elevated-temperature design of Liquid Metal Fast Breeder Reactor (LMFBR) systems and some gas-cooled systems. The U.S. Nuclear Regulatory Commission (NRC) and its Advisory Committee for Reactor Safeguards (ACRS) reviewed the design limits and procedures in the process of reviewing the Clinch River Breeder Reactor (CRBR) for a construction permit in the late 1970s and early 1980s, and identified issues that needed resolution. In the years since then, the NRC and various contractors have evaluated the applicability of the ASME Code and Code Cases to high-temperature reactor designs such as the VHTGRs, and identified issues that need to be resolved to provide a regulatory basis for licensing. This Report describes: (1) NRC and ACRS safety concerns raised during the licensing process of CRBR , (2) how some of these issues are addressed by the current Subsection NH of the ASME Code; and (3) the material models, design criteria, and analysis methods that need to be added to the ASME Code and Code Cases to cover unresolved regulatory issues for very high temperature service.

  13. REGULATORY STRATEGIES TO MINIMIZE GENERATION OF REGULATED WASTES FROM CLEANUP, CONTINUED USE OR DECOMMISSIONING OF NUCLEAR FACILITIES CONTAMINATED WITH POLYCHLORINATED BIPHENYLS (PCBS) - 11198

    SciTech Connect (OSTI)

    Lowry, N.

    2010-11-05T23:59:59.000Z

    Disposal costs for liquid PCB radioactive waste are among the highest of any category of regulated waste. The high cost is driven by the fact that disposal options are extremely limited. Toxic Substances Control Act (TSCA) regulations require most liquids with PCBs at concentration of {ge} 50 parts-per-million to be disposed by incineration or equivalent destructive treatment. Disposal fees can be as high as $200 per gallon. This figure does not include packaging and the cost to transport the waste to the disposal facility, or the waste generator's labor costs for managing the waste prior to shipment. Minimizing the generation of liquid radioactive PCB waste is therefore a significant waste management challenge. PCB spill cleanups often generate large volumes of waste. That is because the removal of PCBs typically requires the liberal use of industrial solvents followed by a thorough rinsing process. In a nuclear facility, the cleanup process may be complicated by the presence of radiation and other occupational hazards. Building design and construction features, e.g., the presence of open grating or trenches, may also complicate cleanup. In addition to the technical challenges associated with spill cleanup, selection of the appropriate regulatory requirements and approach may be challenging. The TSCA regulations include three different sections relating to the cleanup of PCB contamination or spills. EPA has also promulgated a separate guidance policy for fresh PCB spills that is published as Subpart G of 40 CFR 761 although it is not an actual regulation. Applicability is based on the circumstances of each contamination event or situation. Other laws or regulations may also apply. Identification of the allowable regulatory options is important. Effective communication with stakeholders, particularly regulators, is just as important. Depending on the regulatory path that is taken, cleanup may necessitate the generation of large quantities of regulated waste. Allowable options must be evaluated carefully in order to reduce compliance risks, protect personnel, limit potential negative impacts on facility operations, and minimize the generation of wastes subject to TSCA. This paper will identify critical factors in selecting the appropriate TSCA regulatory path in order to minimize the generation of radioactive PCB waste and reduce negative impacts to facilities. The importance of communicating pertinent technical issues with facility staff, regulatory personnel, and subsequently, the public, will be discussed. Key points will be illustrated by examples from five former production reactors at the DOE Savannah River Site. In these reactors a polyurethane sealant was used to seal piping penetrations in the biological shield walls. During the intense neutron bombardment that occurred during reactor operation, the sealant broke down into a thick, viscous material that seeped out of the piping penetrations over adjacent equipment and walls. Some of the walls were painted with a PCB product. PCBs from the paint migrated into the degraded sealant, creating PCB 'spill areas' in some of these facilities. The regulatory cleanup approach selected for each facility was based on its operational status, e.g., active, inactive or undergoing decommissioning. The selected strategies served to greatly minimize the generation of radioactive liquid PCB waste. It is expected that this information would be useful to other DOE sites, DOD facilities, and commercial nuclear facilities constructed prior to the 1979 TSCA ban on most manufacturing and uses of PCBs.

  14. Nuclear materials 1993 annual report. Volume 8, No. 2

    SciTech Connect (OSTI)

    NONE

    1995-05-01T23:59:59.000Z

    This annual report of the US Nuclear Regulatory Commission`s Office for Analysis and Evaluation of Operational Data (AEOD) describes activities conducted during 1993. The report is published in two parts. NUREG-1272, Vol. 8, No. 1, covers power reactors and presents an overview of the operating experience of the nuclear power industry from the NRC perspective, including comments about the trends of some key performance measures. The report also includes the principal findings and issues identified in AEOD studies over the past year and summarizes information from such sources as licensee event reports, diagnostic evaluations, and reports to the NRC`s Operations Center. NUREG-1272, Vol. 8, No. 2, covers nuclear materials and presents a review of the events and concerns during 1993 associated with the use of licensed material in nonreactor applications, such as personnel overexposures and medical misadministrations. Note that the subtitle of No. 2 has been changed from ``Nonreactors`` to ``Nuclear Materials.`` Both reports also contain a discussion of the Incident Investigation Team program and summarize both the Incident Investigation Team and Augmented Inspection Team reports. Each volume contains a list of the AEOD reports issued from 1980 through 1993.

  15. Comparative analysis of United States and French nuclear power plant siting and construction regulatory policies and their economic consequences

    E-Print Network [OSTI]

    Golay, Michael Warren.

    1977-01-01T23:59:59.000Z

    Despite the substantial commitments of time and money which are devoted to the nuclear power plant siting process, the effectiveness of the system in providing a balanced evaluation of the technical, environmental and ...

  16. NRC Licensing Strategy Development for the NGNP

    SciTech Connect (OSTI)

    Mark R. Holbrook; Trevor Cook

    2008-09-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) project will provide the basis for commercialization of a new generation of advanced nuclear plants that utilize hightemperature gas-cooled reactor (HTGR) technology. The inherently safe HTGR design characteristics can be utilized to supply high-temperature process heat, co-generated electricity, and/or hydrogen for a number of industrial applications (e.g., petrochemical processes). Completion of the NGNP will result in a facility that demonstrates the safety and economics of the design, the commercial industrial potential of the technology, and the viability of the licensing strategy.

  17. Application of Microprocessor-Based Equipment in Nuclear Power Plants - Technical Basis for a Qualification Methodology

    SciTech Connect (OSTI)

    Korsah, K.

    2001-08-24T23:59:59.000Z

    This document (1) summarizes the most significant findings of the ''Qualification of Advanced Instrumentation and Control (I&C) Systems'' program initiated by the Nuclear Regulatory Commission (NRC); (2) documents a comparative analysis of U.S. and European qualification standards; and (3) provides recommendations for enhancing regulatory guidance for environmental qualification of microprocessor-based safety-related systems. Safety-related I&C system upgrades of present-day nuclear power plants, as well as I&C systems of Advanced Light-Water Reactors (ALWRs), are expected to make increasing use of microprocessor-based technology. The Nuclear Regulatory Commission (NRC) recognized that the use of such technology may pose environmental qualification challenges different from current, analog-based I&C systems. Hence, it initiated the ''Qualification of Advanced Instrumentation and Control Systems'' program. The objectives of this confirmatory research project are to (1) identify any unique environmental-stress-related failure modes posed by digital technologies and their potential impact on the safety systems and (2) develop the technical basis for regulatory guidance using these findings. Previous findings from this study have been documented in several technical reports. This final report in the series documents a comparative analysis of two environmental qualification standards--Institute of Electrical and Electronics Engineers (IEEE) Std 323-1983 and International Electrotechnical Commission (IEC) 60780 (1998)--and provides recommendations for environmental qualification of microprocessor-based systems based on this analysis as well as on the findings documented in the previous reports. The two standards were chosen for this analysis because IEEE 323 is the standard used in the U.S. for the qualification of safety-related equipment in nuclear power plants, and IEC 60780 is its European counterpart. In addition, the IEC document was published in 1998, and should reflect any new qualification concerns, from the European perspective, with regard to the use of microprocessor-based safety systems in power plants.

  18. License renewal demonstration program: NRC observations and lessons learned

    SciTech Connect (OSTI)

    Prato, R.J.; Kuo, P.T.; Newberry, S.F.

    1996-12-01T23:59:59.000Z

    This report summarizes the Nuclear Regulatory Commission staff`s observations and lessons learned from the five License Renewal Demonstration Program (LRDP) site visits performed by the staff from March 25, 1996, through August 16, 1996. The LRDP was a Nuclear Energy Institute (NEI) program intended to assess the effectiveness of the guidance provided by NEI 95-10, Revision 0, {open_quotes}Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule,{close_quotes} to implement the requirements of Title 10 of the Code of Federal Regulations, Part 54 (10 CFR Part 54), {open_quotes}Requirements for Renewal of Operating Licenses for Nuclear Power Plants.{close_quotes} In general, NEI 95-10 appeared to contain the basic guidance needed for scoping, screening, identifying aging effects, developing aging management programs, and performing time-limited aging analyses. However, inconsistent implementation of this guidance in some areas was an indication that clarification of existing guidance and/or the inclusion-of some new guidance may be needed for applicants to develop a license renewal program that is consistent with the intent of the rule.

  19. Demonstrating Structural Adequacy of Nuclear Power Plant Containment Structures for Beyond Design-Basis Pressure Loadings

    SciTech Connect (OSTI)

    Braverman, J.I.; Morante, R.

    2010-07-18T23:59:59.000Z

    ABSTRACT Demonstrating the structural integrity of U.S. nuclear power plant (NPP) containment structures, for beyond design-basis internal pressure loadings, is necessary to satisfy Nuclear Regulatory Commission (NRC) requirements and performance goals. This paper discusses methods for demonstrating the structural adequacy of the containment for beyond design-basis pressure loadings. Three distinct evaluations are addressed: (1) estimating the ultimate pressure capacity of the containment structure (10 CFR 50 and US NRC Standard Review Plan, Section 3.8) ; (2) demonstrating the structural adequacy of the containment subjected to pressure loadings associated with combustible gas generation (10 CFR 52 and 10 CFR 50); and (3) demonstrating the containment structural integrity for severe accidents (10 CFR 52 as well as SECY 90-016, SECY 93-087, and related NRC staff requirements memoranda (SRMs)). The paper describes the technical basis for specific aspects of the methods presented. It also presents examples of past issues identified in licensing activities related to these evaluations.

  20. 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.

  1. Insights into the behavior of nuclear power plant containments during severe accidents

    SciTech Connect (OSTI)

    Horschel, D.S.; Ludwigsen, J.S.; Parks, M.B.; Lambert, L.D. [Sandia National Labs., Albuquerque, NM (United States); Dameron, R.A.; Rashid, Y.R. [ANATECH Research Corp., San Diego, CA (United States)

    1993-06-01T23:59:59.000Z

    The containment building surrounding a nuclear reactor offers the last barrier to the release of radioactive materials from a severe accident into the environment. The loading environment of the containment under severe accident conditions may include much greater than design pressures and temperatures. Investigations into the performance of containments subject to ultimate or failure pressure and temperature conditions have been performed over the last several years through a program administered by the Nuclear Regulatory Commission (NRC). These NRC sponsored investigations are subsequently discussed. Reviewed are the results of large scale experiments on reinforced concrete, prestressed concrete, and steel containment models pressurized to failure. In conjunction with these major tests, the results of separate effect testing on many of the critical containment components; that is, aged and unaged seals, a personnel air lock and electrical penetration assemblies subjected to elevated temperature and pressure have been performed. An objective of the NRC program is to gain an understanding of the behavior of typical existing and planned containment designs subject to postulated severe accident conditions. This understanding has led to the development of experimentally verified analytical tools that can be applied to accurately predict their ultimate capacities useful in developing severe accident mitigation schemes. Finally, speculation on the response of containments subjected to severe accident conditions is presented.

  2. Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254

    SciTech Connect (OSTI)

    Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

    2012-07-01T23:59:59.000Z

    Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)

  3. Regulatory Guide 5.29, Revision 2, "Special Nuclear Material Control and Accounting Systems for Nuclear Power Plants".

    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: 5-13-14 FEDERALAmerica TreatyWastewantsRequests||NuclearNUCLEAR

  4. Revised analyses of decommissioning for the reference boiling water reactor power station. Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Appendices, draft report for comment. Volume 2

    SciTech Connect (OSTI)

    Smith, R.I.; Bierschbach, M.C.; Konzek, G.J. [Pacific Northwest Lab., Richland, WA (United States)] [and others

    1994-09-01T23:59:59.000Z

    On June 27, 1988, the U.S. Nuclear Regulatory Commission (NRC) published in the Federal Register (53 FR 24018) the final rule for the General Requirements for Decommissioning Nuclear Facilities. With the issuance of the final rule, owners and operators of licensed nuclear power plants are required to prepare, and submit to the NRC for review, decommissioning plans and cost estimates. The NRC staff is in need of updated bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to update the needed bases documentation. This report presents the results of a review and reevaluation of the PNL 1980 decommissioning study of the Washington Public Power Supply System`s WNP-2, including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives, which now include an initial 5-7 year period during which time the spent fuel is stored in the spent fuel pool prior to beginning major disassembly or extended safe storage of the plant. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste. Costs for labor, materials, transport, and disposal activities are given in 1993 dollars. Sensitivities of the total license termination cost to the disposal costs at different low-level radioactive waste disposal sites, to different depths of contaminated concrete surface removal within the facilities, and to different transport distances are also examined.

  5. Strategic Plan for Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

    SciTech Connect (OSTI)

    Kimberlyn C. Mousseau

    2011-10-01T23:59:59.000Z

    The Nuclear Energy Computational Fluid Dynamics Advanced Modeling and Simulation (NE-CAMS) system is being developed at the Idaho National Laboratory (INL) in collaboration with Bettis Laboratory, Sandia National Laboratory (SNL), Argonne National Laboratory (ANL), Utah State University (USU), and other interested parties with the objective of developing and implementing a comprehensive and readily accessible data and information management system for computational fluid dynamics (CFD) verification and validation (V&V) in support of nuclear energy systems design and safety analysis. The two key objectives of the NE-CAMS effort are to identify, collect, assess, store and maintain high resolution and high quality experimental data and related expert knowledge (metadata) for use in CFD V&V assessments specific to the nuclear energy field and to establish a working relationship with the U.S. Nuclear Regulatory Commission (NRC) to develop a CFD V&V database, including benchmark cases, that addresses and supports the associated NRC regulations and policies on the use of CFD analysis. In particular, the NE-CAMS system will support the Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program, which aims to develop and deploy advanced modeling and simulation methods and computational tools for reliable numerical simulation of nuclear reactor systems for design and safety analysis. Primary NE-CAMS Elements There are four primary elements of the NE-CAMS knowledge base designed to support computer modeling and simulation in the nuclear energy arena as listed below. Element 1. The database will contain experimental data that can be used for CFD validation that is relevant to nuclear reactor and plant processes, particularly those important to the nuclear industry and the NRC. Element 2. Qualification standards for data evaluation and classification will be incorporated and applied such that validation data sets will result in well-defined, well-characterized data. Element 3. Standards will be established for the design and operation of experiments for the generation of new validation data sets that are to be submitted to NE-CAMS that addresses the completeness and characterization of the dataset. Element 4. Standards will be developed for performing verification and validation (V&V) to establish confidence levels in CFD analyses of nuclear reactor processes; such processes will be acceptable and recognized by both CFD experts and the NRC.

  6. Structure/piping sensitivity studies for the US NRC Seismic Safety Margins Research Program. [PWR; BWR

    SciTech Connect (OSTI)

    Shieh, L.C.; O'Connell, W.J.; Johnson, J.J.

    1983-01-01T23:59:59.000Z

    The Seismic Safety Margins Research Program (SSMRP) is a NRC-funded, multi-year program conducted by Lawrence Livermore National Laboratory (LLNL). One of the goals of the program is to develop a complete, fully coupled analysis procedure (including methods and computer codes) for estimating the risk of an earthquake-caused radioactive release from commercial nuclear power plant. The analysis procedure is based upon a state-of-the-art evaluation of the current seismic analysis design process and explicitly includes the uncertainties inherent in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. In Phase I, a probabilistic computational procedure was developed for the seismic safety assessment. In Phase II, sensitivity studies were performed, codes and models were improved, and analysis of the Zion plant was completed.

  7. Revised analyses of decommissioning for the reference pressurized Water Reactor Power Station. Volume 2, Effects of current regulatory and other considerations on the financial assurance requirements of the decommissioning rule and on estimates of occupational radiation exposure: Appendices, Final report

    SciTech Connect (OSTI)

    Konzek, G.J.; Smith, R.I.; Bierschbach, M.C.; McDuffie, P.N.

    1995-11-01T23:59:59.000Z

    With the issuance of the final Decommissioning Rule (July 27, 1998), owners and operators of licensed nuclear power plants are required to prepare, and submit to the US Nuclear Regulatory Commission (NRC) for review, decommissioning plans and cost estimates. The NRC staff is in need of bases documentation that will assist them in assessing the adequacy of the licensee submittals, from the viewpoint of both the planned actions, including occupational radiation exposure, and the probable costs. The purpose of this reevaluation study is to provide some of the needed bases documentation. This report contains the results of a review and reevaluation of the 1978 PNL decommissioning study of the Trojan nuclear power plant (NUREG/CR-0130), including all identifiable factors and cost assumptions which contribute significantly to the total cost of decommissioning the nuclear power plant for the DECON, SAFSTOR, and ENTOMB decommissioning alternatives. These alternatives now include an initial 5--7 year period during which time the spent fuel is stored in the spent fuel pool, prior to beginning major disassembly or extended safe storage of the plant. Included for information (but not presently part of the license termination cost) is an estimate of the cost to demolish the decontaminated and clean structures on the site and to restore the site to a ``green field`` condition. This report also includes consideration of the NRC requirement that decontamination and decommissioning activities leading to termination of the nuclear license be completed within 60 years of final reactor shutdown, consideration of packaging and disposal requirements for materials whose radionuclide concentrations exceed the limits for Class C low-level waste (i.e., Greater-Than-Class C), and reflects 1993 costs for labor, materials, transport, and disposal activities.

  8. 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.

  9. SSI sensitivity studies and model improvements for the US NRC Seismic Safety Margins Research Program. Rev. 1

    SciTech Connect (OSTI)

    Johnson, J.J.; Maslenikov, O.R.; Benda, B.J.

    1984-10-01T23:59:59.000Z

    The Seismic Safety Margins Research Program (SSMRP) is a US NRC-funded program conducted by Lawrence Livermore National Laboratory. Its goal is to develop a complete fully coupled analysis procedure for estimating the risk of an earthquake-induced radioactive release from a commercial nuclear power plant. In Phase II of the SSMRP, the methodology was applied to the Zion nuclear power plant. Three topics in the SSI analysis of Zion were investigated and reported here - flexible foundation modeling, structure-to-structure interaction, and basemat uplift. The results of these investigations were incorporated in the SSMRP seismic risk analysis. 14 references, 51 figures, 13 tables.

  10. Survey of ambient electromagnetic and radio-frequency interference levels in nuclear power plants

    SciTech Connect (OSTI)

    Kercel, S.W.; Moore, M.R.; Blakeman, E.D.; Ewing, P.D.; Wood, R.T.

    1996-11-01T23:59:59.000Z

    This document reports the results of a survey of ambient electromagnetic conditions in representative nuclear power plants. The U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research engaged the Oak Ridge National Laboratory (ORNL) to perform these measurements to characterize the electromagnetic interference (EMI) and radio-frequency interference (RFI) levels that can be expected in nuclear power plant environments. This survey is the first of its kind, being based on long-term unattended observations. The data presented in this report were measured at eight different nuclear units and required 14 months to collect. A representative sampling of power plant conditions (reactor type, operating mode, site location) monitored over extended observation periods (up to 5 weeks) were selected to more completely determine the characteristic electromagnetic environment for nuclear power plants. Radiated electric fields were measured over the frequency range of 5 MHz to 8 GHz. Radiated magnetic fields and conducted EMI events were measured over the frequency range of 305 Hz to 5 MHz. Highest strength observations of the electromagnetic ambient environment across all measurement conditions at each site provide frequency-dependent profiles for EMI/RFI levels in nuclear power plants.

  11. Standard Guide for Preparing Waste Management Plans for Decommissioning Nuclear Facilities

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 This guide addresses the development of waste management plans for potential waste streams resulting from decommissioning activities at nuclear facilities, including identifying, categorizing, and handling the waste from generation to final disposal. 1.2 This guide is applicable to potential waste streams anticipated from decommissioning activities of nuclear facilities whose operations were governed by the Nuclear Regulatory Commission (NRC) or Agreement State license, under Department of Energy (DOE) Orders, or Department of Defense (DoD) regulations. 1.3 This guide provides a description of the key elements of waste management plans that if followed will successfully allow for the characterization, packaging, transportation, and off-site treatment or disposal, or both, of conventional, hazardous, and radioactive waste streams. 1.4 This guide does not address the on-site treatment, long term storage, or on-site disposal of these potential waste streams. 1.5 This standard does not purport to address ...

  12. Early Site Permit Demonstration Program: Regulatory criteria evaluation report

    SciTech Connect (OSTI)

    Not Available

    1993-03-01T23:59:59.000Z

    The primary objective of the ESPDP is to demonstrate successfully the use of 10CFR52 to obtain ESPs for one or more US sites for one (or more) ALWR nuclear power plants. It is anticipated that preparation of the ESP application and interaction with NRC during the application review process will result not only in an ESP for the applicant(s) but also in the development of criteria and definition of processes, setting the precedent that facilitates ESPs for subsequent ESP applications. Because siting regulatory processes and acceptance criteria are contained in over 100 separate documents, comprehensive licensing and technical reviews were performed to establish whether the requirements and documentation are self-consistent, whether the acceptance criteria are sufficiently well-defined and clear, and whether the licensing process leading to the issuance of an ESP is unambiguously specified. The results of the technical and licensing evaluations are presented in this report. The purpose, background, and organization of the ESPDP is delineated in Section 1. Section 11 contains flowcharts defining siting application requirements, environmental report requirements, and emergency planning/preparedness requirements for ALWRS. The licensing and technical review results are presented in Section III.

  13. Using micro saint to predict performance in a nuclear power plant control room

    SciTech Connect (OSTI)

    Lawless, M.T.; Laughery, K.R. [Micro Analysis and Design, Inc., Boulder, CO (United States); Persenky, J.J. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-09-01T23:59:59.000Z

    The United States Nuclear Regulatory Commission (NRC) requires a technical basis for regulatory actions. In the area of human factors, one possible technical basis is human performance modeling technology including task network modeling. This study assessed the feasibility and validity of task network modeling to predict the performance of control room crews. Task network models were built that matched the experimental conditions of a study on computerized procedures that was conducted at North Carolina State University. The data from the {open_quotes}paper procedures{close_quotes} conditions were used to calibrate the task network models. Then, the models were manipulated to reflect expected changes when computerized procedures were used. These models` predictions were then compared to the experimental data from the {open_quotes}computerized conditions{close_quotes} of the North Carolina State University study. Analyses indicated that the models predicted some subsets of the data well, but not all. Implications for the use of task network modeling are discussed.

  14. Nuclear containment steel liner corrosion workshop : final summary and recommendation report.

    SciTech Connect (OSTI)

    Erler, Bryan A. (Erler Engineering Ltd., Chicago, IL); Weyers, Richard E. (Virginia Tech University, Blacksburg, VA); Sagues, Alberto (University of South Florida, Tampa, FL); Petti, Jason P.; Berke, Neal Steven (Tourney Consulting Group, LLC, Kalamazoo, MI); Naus, Dan J. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2011-07-01T23:59:59.000Z

    This report documents the proceedings of an expert panel workshop conducted to evaluate the mechanisms of corrosion for the steel liner in nuclear containment buildings. The U.S. Nuclear Regulatory Commission (NRC) sponsored this work which was conducted by Sandia National Laboratories. A workshop was conducted at the NRC Headquarters in Rockville, Maryland on September 2 and 3, 2010. Due to the safety function performed by the liner, the expert panel was assembled in order to address the full range of issues that may contribute to liner corrosion. This report is focused on corrosion that initiates from the outer surface of the liner, the surface that is in contact with the concrete containment building wall. Liner corrosion initiating on the outer diameter (OD) surface has been identified at several nuclear power plants, always associated with foreign material left embedded in the concrete. The potential contributing factors to liner corrosion were broken into five areas for discussion during the workshop. Those include nuclear power plant design and operation, corrosion of steel in contact with concrete, concrete aging and degradation, concrete/steel non-destructive examination (NDE), and concrete repair and corrosion mitigation. This report also includes the expert panel member's recommendations for future research.

  15. Auxiliary feedwater system risk-based inspection guide for the Point Beach nuclear power plant

    SciTech Connect (OSTI)

    Lloyd, R C; Moffitt, N E; Gore, B F; Vo, T V; Vehec, T A [Pacific Northwest Lab., Richland, WA (United States)

    1993-02-01T23:59:59.000Z

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Point Beach was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRS. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Point Beach plant.

  16. Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant

    SciTech Connect (OSTI)

    Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E. (Pacific Northwest Lab., Richland, WA (United States))

    1991-09-01T23:59:59.000Z

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Ginna was selected as the eighth plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Ginna plant. 23 refs., 1 fig., 1 tab.

  17. Human-reliability data bank for nuclear-power-plant operations. Volume 2. A data-bank concept and system description

    SciTech Connect (OSTI)

    Comer, M.K.; Kozinsky, E.J.; Eckel, J.S.; Miller, D.P.

    1983-02-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) is conducting a research program to determine the need for a human-reliability data bank unique to the nuclear industry. This report, in describing a proposed data bank, focuses primarily on four requirements: human-performance data collection, treatment, structuring (storage), and retrieval. Four data-collection methods are proposed: three are extensions to existing systems (i.e., Licensee Event Reports (LER), Nuclear Plant Reliability Data System (NPRDS), and Plant Incident Reports (PIR)); the fourth is a new system called the Nuclear Safety Reporting System (NSRS). Data treatment involves evaluating raw field data and data from other sources (e.g., training simulator, expert judgment, and performance modeling), and preparing them for entry into the data bank. Data structuring involves storage of data by equipment characteristics and human actions at the system, component, and individual control/display levels. Data retrieval uses a set of matrices based on the data-structuring taxonomy.

  18. Regulatory Framework for Salt Waste Disposal and Tank Closure at the Savannah River Site - 13663

    SciTech Connect (OSTI)

    Thomas, Steve; Dickert, Ginger [Savannah River Remediation LLC, Savannah River Site, Aiken, SC 29808 (United States)] [Savannah River Remediation LLC, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01T23:59:59.000Z

    The end of the Cold War has left a legacy of approximately 37 million gallons of radioactive waste in the aging waste tanks at the Department of Energy's Savannah River Site (SRS). A robust program is in place to remove waste from these tanks, treat the waste to separate into a relatively small volume of high-level waste and a large volume of low-level waste, and to actively dispose of the low-level waste on-site and close the waste tanks and associated ancillary structures. To support performance-based, risk-informed decision making and to ensure compliance with all regulatory requirements, the U.S. Department of Energy (DOE) and its current and past contractors have worked closely with the South Carolina Department of Health and Environmental Control (SCDHEC), the U.S. Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC) to develop and implement a framework for on-site low-level waste disposal and closure of the SRS waste tanks. The Atomic Energy Act of 1954, as amended, provides DOE the authority to manage defense-related radioactive waste. DOE Order 435.1 and its associated manual and guidance documents detail this radioactive waste management process. The DOE also has a requirement to consult with the NRC in determining that waste that formerly was classified as high-level waste can be safely managed as either low-level waste or transuranic waste. Once DOE makes a determination, NRC then has a responsibility to monitor DOE's actions in coordination with SCDHEC to ensure compliance with the Title 10 Code of Federal Regulations Part 61 (10CFR61), Subpart C performance objectives. The management of hazardous waste substances or components at SRS is regulated by SCDHEC and the EPA. The foundation for the interactions between DOE, SCDHEC and EPA is the SRS Federal Facility Agreement (FFA). Managing this array of requirements and successfully interacting with regulators, consultants and stakeholders is a challenging task but ensures thorough and thoughtful processes for disposing of the SRS low-level waste and the closure of the tank farm facilities. (authors)

  19. Full-Scale Accident Testing in Support of Used Nuclear Fuel Transportation.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Lindgren, Eric R.; Rechard, Rob P.; Sorenson, Ken B.

    2014-09-01T23:59:59.000Z

    The safe transport of spent nuclear fuel and high-level radioactive waste is an important aspect of the waste management system of the United States. The Nuclear Regulatory Commission (NRC) currently certifies spent nuclear fuel rail cask designs based primarily on numerical modeling of hypothetical accident conditions augmented with some small scale testing. However, NRC initiated a Package Performance Study (PPS) in 2001 to examine the response of full-scale rail casks in extreme transportation accidents. The objectives of PPS were to demonstrate the safety of transportation casks and to provide high-fidelity data for validating the modeling. Although work on the PPS eventually stopped, the Blue Ribbon Commission on America’s Nuclear Future recommended in 2012 that the test plans be re-examined. This recommendation was in recognition of substantial public feedback calling for a full-scale severe accident test of a rail cask to verify evaluations by NRC, which find that risk from the transport of spent fuel in certified casks is extremely low. This report, which serves as the re-assessment, provides a summary of the history of the PPS planning, identifies the objectives and technical issues that drove the scope of the PPS, and presents a possible path for moving forward in planning to conduct a full-scale cask test. Because full-scale testing is expensive, the value of such testing on public perceptions and public acceptance is important. Consequently, the path forward starts with a public perception component followed by two additional components: accident simulation and first responder training. The proposed path forward presents a series of study options with several points where the package performance study could be redirected if warranted.

  20. 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,...

  1. Cleanup Contractor Achieves 'Elite' Nuclear Material Accountability...

    Energy Savers [EERE]

    the possession, use, and shipment of nuclear material within the U.S., as well as all exports and imports of such material. The database is jointly funded by the NRC and DOE and...

  2. Independent Verification and Validation Of SAPHIRE 8 Software Design and Interface Design Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Kent Norris

    2009-10-01T23:59:59.000Z

    The purpose of the Independent Verification and Validation (IV&V) role in the evaluation of the SAPHIRE software design and interface design is to assess the activities that results in the development, documentation, and review of a software design that meets the requirements defined in the software requirements documentation. The IV&V team began this endeavor after the software engineering and software development of SAPHIRE had already been in production. IV&V reviewed the requirements specified in the NRC Form 189s to verify these requirements were included in SAPHIRE’s Software Verification and Validation Plan (SVVP) design specification.

  3. Independent Verification and Validation Of SAPHIRE 8 Software Design and Interface Design Project Number: N6423 U.S. Nuclear Regulatory Commission

    SciTech Connect (OSTI)

    Kent Norris

    2010-03-01T23:59:59.000Z

    The purpose of the Independent Verification and Validation (IV&V) role in the evaluation of the SAPHIRE software design and interface design is to assess the activities that results in the development, documentation, and review of a software design that meets the requirements defined in the software requirements documentation. The IV&V team began this endeavor after the software engineering and software development of SAPHIRE had already been in production. IV&V reviewed the requirements specified in the NRC Form 189s to verify these requirements were included in SAPHIRE’s Software Verification and Validation Plan (SVVP) design specification.

  4. A Comparison of International Regulatory Organizations and

    E-Print Network [OSTI]

    Aickelin, Uwe

    the safety regulation and the licensing of new nuclear power plants. The paper considers both design safety great emphasis to safety issues during the licensing process of any new nuclear plant. FurthermoreA Comparison of International Regulatory Organizations and Licensing Procedures for New Nuclear

  5. Proceedings of the U. S. Nuclear Regulatory Commission review group conference on advanced instrumentation research for reactor safety held at Oak Ridge National Laboratory on July 29-31, 1980. Conference proceedings

    SciTech Connect (OSTI)

    Hon, A.L.; Basdekas, D.; Hsu, Y.Y.; Kondic, N.; Van Houten, R.

    1980-12-01T23:59:59.000Z

    The report compiles the technical presentations during the Advanced Instrumentation Research for the Reactor Safety Review Group Meeting held in July 1980. The three-day meeting covered the Power Plant Instrumentation, Two-Phase Flow Instrumentation, Fuel Behavior Research Instrumentation and Advanced Reactor Instrumentation research programs sponsored by the U.S. Nuclear Regulatory Commission, Division of Reactor Safety Research. In addition, two invited papers from the nuclear industry were also presented. The conference is held each year to review the up-to-date instrumentation research results by the contractors. It also provides the opportunity for the researchers and experts to exchange experience on advanced instrumentation development. The report serves as a vehicle to disseminate the state-of-the-art information to the research community and the nuclear industry.

  6. Independent Confirmatory Survey Report for the University of Arizona Nuclear Reactor Laboratory, Tucson, Arizona

    SciTech Connect (OSTI)

    Nick A. Altic

    2011-11-11T23:59:59.000Z

    The University of Arizona (University) research reactor is a TRIGA swimming pool type reactor designed by General Atomics and constructed at the University in 1958. The reactor first went into operation in December of 1958 under U.S. Nuclear Regulatory Commission (NRC) license R-52 until final shut down on May 18, 2010. Initial site characterization activities were conducted in February 2009 during ongoing reactor operations to assess the radiological status of the Nuclear Reactor Laboratory (NRL) excluding the reactor tank, associated components, and operating systems. Additional post-shutdown characterization activities were performed to complete characterization activities as well as verify assumptions made in the Decommissioning Plan (DP) that were based on a separate activation analysis (ESI 2009 and WMG 2009). Final status survey (FSS) activities began shortly after the issuance of the FSS plan in May 2011. The contractor completed measurement and sampling activities during the week of August 29, 2011.

  7. Knowledges and abilities catalog for nuclear power plant operators: pressurized water reactors

    SciTech Connect (OSTI)

    Not Available

    1985-07-01T23:59:59.000Z

    This document catalogs roughly 5300 knowledges and abilities of reactor operators and senior reactor operators. It results from a reanalysis of much larger job-task analysis data base compiled by the Institute of Nuclear Power Operations (INPO). Knowledges and abilities are cataloged for 45 major power plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functions (e.g., reactivity control and reactor coolant system inventory control). With appropriate sampling from this catalog, operator licensing examinations having content validity can be developed. A structured sampling procedure for this catalog is under development by the Nuclear Regulatory Commission (NRC) and will be published as a companion document, ''Examiners' Handbook for Developing Operator Licensing Examinations'' (NUREG-1121). The examinations developed by using the catalog and handbook will cover those topics listed under Title 10, Code of Federal Regulations, Part 55.

  8. 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.

  9. Recent MELCOR and VICTORIA Fission Product Research at the NRC

    SciTech Connect (OSTI)

    Bixler, N.E.; Cole, R.K.; Gauntt, R.O.; Schaperow, J.H.; Young, M.F.

    1999-01-21T23:59:59.000Z

    The MELCOR and VICTORIA severe accident analysis codes, which were developed at Sandia National Laboratories for the U. S. Nuclear Regulatory Commission, are designed to estimate fission product releases during nuclear reactor accidents in light water reactors. MELCOR is an integrated plant-assessment code that models the key phenomena in adequate detail for risk-assessment purposes. VICTORIA is a more specialized fission- product code that provides detailed modeling of chemical reactions and aerosol processes under the high-temperature conditions encountered in the reactor coolant system during a severe reactor accident. This paper focuses on recent enhancements and assessments of the two codes in the area of fission product chemistry modeling. Recently, a model for iodine chemistry in aqueous pools in the containment building was incorporated into the MELCOR code. The model calculates dissolution of iodine into the pool and releases of organic and inorganic iodine vapors from the pool into the containment atmosphere. The main purpose of this model is to evaluate the effect of long-term revolatilization of dissolved iodine. Inputs to the model include dose rate in the pool, the amount of chloride-containing polymer, such as Hypalon, and the amount of buffering agents in the containment. Model predictions are compared against the Radioiodine Test Facility (RTF) experiments conduced by Atomic Energy of Canada Limited (AECL), specifically International Standard Problem 41. Improvements to VICTORIA's chemical reactions models were implemented as a result of recommendations from a peer review of VICTORIA that was completed last year. Specifically, an option is now included to model aerosols and deposited fission products as three condensed phases in addition to the original option of a single condensed phase. The three-condensed-phase model results in somewhat higher predicted fission product volatilities than does the single-condensed-phase model. Modeling of U02 thermochemistry was also improved, and results in better prediction of vaporization of uranium from fuel, which can react with released fission products to affect their volatility. This model also improves the prediction of fission product release rates from fuel. Finally, recent comparisons of MELCOR and VICTORIA with International Standard Problem 40 (STORM) data are presented. These comparisons focus on predicted therrnophoretic deposition, which is the dominant deposition mechanism. Sensitivity studies were performed with the codes to examine experimental and modeling uncertainties.

  10. NEXT GENERATION NUCLEAR PLANT LICENSING BASIS EVENT SELECTION WHITE PAPER

    SciTech Connect (OSTI)

    Mark Holbrook

    2010-09-01T23:59:59.000Z

    The Next Generation Nuclear Plant (NGNP) will be a licensed commercial high temperature gas-cooled reactor (HTGR) plant capable of producing the electricity and high temperature process heat for industrial markets supporting a range of end-user applications. The NGNP Project has adopted the 10 CFR 52 Combined License (COL) application process, as recommended in the Report to Congress, dated August 2008, as the foundation for the NGNP licensing strategy. NRC licensing of the NGNP plant utilizing this process will demonstrate the efficacy of licensing future HTGRs for commercial industrial applications. This white paper is one in a series of submittals that will address key generic issues of the COL priority licensing topics as part of the process for establishing HTGR regulatory requirements.

  11. UNITED STATES NUCLEAR REGULATORY COMMISSION

    Office of Legacy Management (LM)

    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 group currentBradleyTableSelling7 AugustAFRICAN3uj: ;;IDEC. i' ,' ,'

  12. Regulatory Tools

    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 PossibleRadiation Protection Technical s o Freiberg and SondershausenRegulatory

  13. ISDAC - NRC Convair-580 Flight Hours Date Flight

    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 PowerCherries 82981-1cnHigh School footballHydrogen andHypernucleiNORTHWESTOffice ofOffice- NRC

  14. DOE/NRC F 740M | Department of Energy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742Energy China 2015 Business42.1 DEPARTMENTSeptember 27,September 24, 2002The RoundtableDOE/NRC

  15. NRC Institute for Fuel Cell Innovation | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov YouKizildere I Geothermal Pwer PlantMunhall, Pennsylvania: Energy Resources JumpNEFAppropriationReference Manual 53:NRC

  16. NEW MATERIALS DEVELOPED TO MEET REGULATORY AND TECHNICAL REQUIREMENTS ASSOCIATED WITH IN-SITU DECOMMISSIONING OF NUCLEAR REACTORS AND ASSOCIATED FACILITIES

    SciTech Connect (OSTI)

    Blankenship, J.; Langton, C.; Musall, J.; Griffin, W.

    2012-01-18T23:59:59.000Z

    For the 2010 ANS Embedded Topical Meeting on Decommissioning, Decontamination and Reutilization and Technology, Savannah River National Laboratory's Mike Serrato reported initial information on the newly developed specialty grout materials necessary to satisfy all requirements associated with in-situ decommissioning of P-Reactor and R-Reactor at the U.S. Department of Energy's Savannah River Site. Since that report, both projects have been successfully completed and extensive test data on both fresh properties and cured properties has been gathered and analyzed for a total of almost 191,150 m{sup 3} (250,000 yd{sup 3}) of new materials placed. The focus of this paper is to describe the (1) special grout mix for filling the P-Reactor vessel (RV) and (2) the new flowable structural fill materials used to fill the below grade portions of the facilities. With a wealth of data now in hand, this paper also captures the test results and reports on the performance of these new materials. Both reactors were constructed and entered service in the early 1950s, producing weapons grade materials for the nation's defense nuclear program. R-Reactor was shut down in 1964 and the P-Reactor in 1991. In-situ decommissioning (ISD) was selected for both facilities and performed as Comprehensive Environmental Response, Compensations and Liability Act actions (an early action for P-Reactor and a removal action for R-Reactor), beginning in October 2009. The U.S. Department of Energy concept for ISD is to physically stabilize and isolate intact, structurally robust facilities that are no longer needed for their original purpose of producing (reactor facilities), processing (isotope separation facilities), or storing radioactive materials. Funding for accelerated decommissioning was provided under the American Recovery and Reinvestment Act. Decommissioning of both facilities was completed in September 2011. ISD objectives for these CERCLA actions included: (1) Prevent industrial worker exposure to radioactive or hazardous contamination exceeding Principal Threat Source Material levels; (2) Minimize human and ecological exposure to unacceptable risk associated with radiological and hazardous constituents that are or may be present; (3) Prevent to the extent practicable the migration of radioactive or hazardous contaminants from the closed facility to the groundwater so that concentrations in groundwater do not exceed regulatory standards; (4) Eliminate or control all routes of human exposure to radiological and chemical contamination; and (5) Prevent animal intruder exposure to radioactive and hazardous contamination.

  17. 5.0 SUMMARY

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

    NRC (U.S. Nuclear Regulatory Commission) . Regulatory Guide 1.109. RAPCA (Regional Air Pollution Control Agency) . 1988. Air Quality Report Air Monitoring Data for 1986,...

  18. Microsoft PowerPoint - 2_Peter J. Habighorst_NRC Remarks 2013...

    National Nuclear Security Administration (NNSA)

    General licenses 8 NRC Internal actions, cont. * Exports to power reactors in Canada, Germany and Japan were identified: - KNK, THTR, AVR, JOYO, FUGEN, Bruce * General licenses...

  19. Microsoft PowerPoint - 8_Peter Habighorst_NRC_Act of 2012-status...

    National Nuclear Security Administration (NNSA)

    General licenses 8 NRC Internal actions, cont. * Exports to power reactors in Canada, Germany and Japan were identified: - KNK, THTR, AVR, JOYO, FUGEN, Bruce * General licenses...

  20. Regulatory Perspective on Potential Fuel Reconfiguration and Its Implication to High Burnup Spent Fuel Storage and Transportation - 13042

    SciTech Connect (OSTI)

    Li, Zhian; Rahimi, Meraj; Tang, David; Aissa, Mourad; Flaganan, Michelle [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States)] [U.S. Nuclear Regulatory Commission - NRC, Washington, DC 20555-0001 (United States); Wagner, John C. [Oak Ridge National Laboratory (United States)] [Oak Ridge National Laboratory (United States)

    2013-07-01T23:59:59.000Z

    The recent experiments conducted by Argonne National Laboratory on high burnup fuel cladding material property show that the ductile to brittle transition temperature of high burnup fuel cladding is dependent on: (1) cladding material, (2) irradiation conditions, and (3) drying-storage histories (stress at maximum temperature) [1]. The experiment results also show that the ductile to brittle temperature increases as the fuel burnup increases. These results indicate that the current knowledge in cladding material property is insufficient to determine the structural performance of the cladding of high burnup fuel after it has been stored in a dry cask storage system for some time. The uncertainties in material property and the elevated ductile to brittle transition temperature impose a challenge to the storage cask and transportation packaging designs because the cask designs may not be able to rely on the structural integrity of the fuel assembly for control of fissile material, radiation source, and decay heat source distributions. The fuel may reconfigure during further storage and/or the subsequent transportation conditions. In addition, the fraction of radioactive materials available for release from spent fuel under normal condition of storage and transport may also change. The spent fuel storage and/or transportation packaging vendors, spent fuel shippers, and the regulator may need to consider this possible fuel reconfiguration and its impact on the packages' ability to meet the safety requirements of Part 72 and Part 71 of Title 10 of the Code of Federal Regulations. The United States Nuclear Regulatory Commission (NRC) is working with the scientists at Oak Ridge National Laboratory (ORNL) to assess the impact of fuel reconfiguration on the safety of the dry storage systems and transportation packages. The NRC Division of Spent Fuel Storage and Transportation has formed a task force to work on the safety and regulatory concerns in relevance to high burnup fuel storage and transportation. This paper discusses the staff's preliminary considerations on the safety implication of fuel reconfiguration with respect to nuclear safety (subcriticality control), radiation shielding, containment, the performance of the thermal functions of the packages, and the retrievability of the contents from regulatory perspective. (authors)

  1. Top U.S. Nuclear Official Commends Industry for Submitting 3rd...

    Energy Savers [EERE]

    Top U.S. Nuclear Official Commends Industry for Submitting 3rd Combined Construction & Operating License Application to the NRC Top U.S. Nuclear Official Commends Industry for...

  2. Power to the People or Regulatory Ratcheting? Explaining the Success (or Failure) of Attempts to Site Commercial U.S. Nuclear Power Plants: 1954 -19961

    E-Print Network [OSTI]

    to Site Commercial U.S. Nuclear Power Plants: 1954 - 19961 7 April 2014 Eric Berndt2 and Daniel P. Aldrich to attempt siting nuclear power plant facilities in large numbers in the 1960s. By the late 1990s, more than 1984). In the case of the Shoreham Nuclear Generating Station in Long Island, the plant was completed

  3. A Review of Information for Managing Aging in Nuclear Power Plants

    SciTech Connect (OSTI)

    WC Morgan; JV Livingston

    1995-09-01T23:59:59.000Z

    Age related degradation effects in safety related systems of nuclear power plants should be managed to prevent safety margins from eroding below the acceptable limits provided in plant design bases. The Nuclear Plant Aging Research (NPAR) Pro- gram, conducted under the auspices of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, and other related aging management programs are developing technical information on managing aging. The aging management process central to these efforts consists of three key elements: 1) selecting structures, systems, and components (SSCs) in which aging should be controlled; 2) understanding the mechanisms and rates of degradation in these SSCs; and 3) managing degradation through effective inspection, surveillance, condition monitoring, trending, record keeping, mainten- ance, refurbishment, replacement, and adjustments in the operating environment and service conditions. This document concisely reviews and integrates information developed under the NPAR Program and other aging management studies and other available information related to understanding and managing age-related degradation effects and provides specific refer- ences to more comprehensive information on the same subjects.

  4. COMPARISON OF RESULTS FOR QUARTER 5 SURFACE WATER SPLIT SAMPLES COLLECTED AT THE NUCLEAR FUEL SERVICES SITE ERWIN TENNESSEE

    SciTech Connect (OSTI)

    none,

    2013-09-23T23:59:59.000Z

    Oak Ridge Associated Universities (ORAU), under the Oak Ridge Institute for Science and Education (ORISE) contract, collected split surface water samples with Nuclear Fuel Services (NFS) representatives on August 21, 2013. Representatives from the U.S. Nuclear Regulatory Commission (NRC) and the Tennessee Department of Environment and Conservation were also in attendance. Samples were collected at four surface water stations, as required in the approved Request for Technical Assistance number 11-018. These stations included Nolichucky River upstream (NRU), Nolichucky River downstream (NRD), Martin Creek upstream (MCU), and Martin Creek downstream (MCD). Both ORAU and NFS performed gross alpha and gross beta analyses, and the comparison of results using the duplicate error ratio (DER), also known as the normalized absolute difference, are tabulated. All DER values were less than 3 and results are consistent with low (e.g., background) concentrations.

  5. Nuclear power plant Generic Aging Lessons Learned (GALL). Main report and appendix A

    SciTech Connect (OSTI)

    Kaza, K.E.; Diercks, D.R.; Holland, J.W.; Choi, S.U. [and others

    1996-12-01T23:59:59.000Z

    The purpose of this generic aging lessons learned (GALL) review is to provide a systematic review of plant aging information in order to assess materials and component aging issues related to continued operation and license renewal of operating reactors. Literature on mechanical, structural, and thermal-hydraulic components and systems reviewed consisted of 97 Nuclear Plant Aging Research (NPAR) reports, 23 NRC Generic Letters, 154 Information Notices, 29 Licensee Event Reports (LERs), 4 Bulletins, and 9 Nuclear Management and Resources Council Industry Reports (NUMARC IRs) and literature on electrical components and systems reviewed consisted of 66 NPAR reports, 8 NRC Generic Letters, 111 Information Notices, 53 LERs, 1 Bulletin, and 1 NUMARC IR. More than 550 documents were reviewed. The results of these reviews were systematized using a standardized GALL tabular format and standardized definitions of aging-related degradation mechanisms and effects. The tables are included in volume s 1 and 2 of this report. A computerized data base has also been developed for all review tables and can be used to expedite the search for desired information on structures, components, and relevant aging effects. A survey of the GALL tables reveals that all ongoing significant component aging issues are currently being addressed by the regulatory process. However, the aging of what are termed passive components has been highlighted for continued scrutiny. This document is Volume 1, consisting of the executive summary, summary and observations, and an appendix listing the GALL literature review tables.

  6. NRC, SPI, and chasing arrows: Is there common ground

    SciTech Connect (OSTI)

    Rabasca, L.

    1994-12-01T23:59:59.000Z

    After negotiating for 15 months, issuing three white papers, and conducting consumer research, the National Recycling Coalition (NRC, Washington, DC) and the Society of the Plastics Industry, Inc. (SPI, Washington, DC), have agreed to disagree on the use of the familiar chasing arrows logo in SPI's seven-number resin identification code. The desired end result of the talks and debates was supposed to be a plan to change legislation requiring 39 states to use SPI's current resin identification code and a commitment to remove the old code from durable goods and flexible packaging. Ultimately, these actions could have improved markets for polyethylene terephthalate (PET) and high-density polyethylene (HDPE) by reducing contamination caused by confusion over what is actually recycled versus what is merely recyclable.

  7. Risk Informing the Commercial Nuclear Enterprise

    E-Print Network [OSTI]

    Bernstein, Joseph B.

    % Moderate 50% VeryLikely 95% Likely 80% Remote 20% CriticalInsignificant Minor Significant Major Impact Post-Fukushima Response New NRC Regulations EPA Cooling Water Intake regulation GSI 191 4 Cyber and MD sites Regulatory Critical Remote Low Industry proposing alternatives to federal and state EPA Key

  8. Nuclear reactors built, being built, or planned, 1994

    SciTech Connect (OSTI)

    NONE

    1995-07-01T23:59:59.000Z

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  9. Nuclear reactors built, being built, or planned: 1995

    SciTech Connect (OSTI)

    NONE

    1996-08-01T23:59:59.000Z

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  10. Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

    SciTech Connect (OSTI)

    P.M. O'Leary; Dr. M.L. Pitts

    2000-08-21T23:59:59.000Z

    Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers.

  11. A PROPOSED METHOD FOR ESTIMATING FAILURE RATES OF DEGRADED PASSIVE COMPONENTS IN THE NRC SIGNIFICANCE DETERMINATION PROCESS

    SciTech Connect (OSTI)

    Unwin, Stephen D.; Johnson, Kenneth I.; Ivans, William J.; Lowry, Peter P.

    2013-11-01T23:59:59.000Z

    This paper outlines a methodology for estimation of the incremental core damage frequency associated with the degradation of passive components with a view to its application in the US Nuclear Regulatory Commission's significance determination process. The method involves use of simplified physics-based models of materials degradation and the probabilistic comparison of transient loads with deteriorating system capacities.

  12. UnitedStates EnvironmentalProtection

    E-Print Network [OSTI]

    ),theDOEOfficeofEnvironmentalRes- toration and WasteManagement(EM),and the NuclearRegulatoryCommission(NRC)Officeof Nuclear radioactivematerialsandsitesin theNRC'sSite Decommissioning Management Program (SDMP). Contents of Report Thereport includesan

  13. Modeling prismatic HTGRs with U.S. N.R.C advanced gas reactor evaluator (AGREE)

    SciTech Connect (OSTI)

    Seker, V.; Drzewiecki, T.; Downar, T. [Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Blvd, Ann Arbor, MI 48109 (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington, DC (United States)

    2012-07-01T23:59:59.000Z

    A core fluids and heat transfer model has been developed for the prismatic high temperature gas reactor in support of the US NRC Next Generation Nuclear Plant (NGNP) evaluation model. The core fluids modeling relies on a subchannel approach in which the primary coolant flow path through the core region and vertical in-core and ex-core gaps can be modeled as individual subchannels. These subchannels are connected together to represent a three dimensional reactor. An initial validation calculation for the core fluids model has been performed using data available in literature for bypass flow. The predicted bypass flow was within 2.6% of the value reported in the literature. The core level heat transfer model is based on a triangular finite volume method, where the base triangle is one sixth of the prismatic block. In order to improve the spatial accuracy at this level, a triangular refinement method was also implemented. The fuel compact temperature is calculated by a cylindrical conduction model which is implicitly coupled to the triangular core level model. The preliminary verification of the model was performed by comparing AGREE to a finite element code COMSOL by analyzing the MHTGR core heat transfer. Further verification and validation is currently an ongoing effort. (authors)

  14. Systems interaction results from the digraph matrix analysis of the Watts Bar Nuclear Power Plant high pressure safety injection systems. Volume 2

    SciTech Connect (OSTI)

    Sacks, I.J.; Ashmore, B.C.; Alesso, H.P.

    1983-07-01T23:59:59.000Z

    Spatial and functional coupling of nuclear power plant systems that lead to interdependencies are called Systems Interactions. At present, the US Nuclear Regulatory Commission (NRC) is investigating ways of integrating a systems interactions study with existing Probabilistic Risk Assessment efforts. One approach is based on graph-theoretic methods utilizing matrix representations of logic diagrams called Digraph Matrix Analysis (DMA). The objective in this report is to demonstrate the capabilities of Digraph Matrix Analysis to model an accident sequence (including front line systems, support systems and human actions) as a continuous, well-integrated logic model in order to identify and evaluate functional systems interactions. The selected accident sequence, loss of high pressure safety injection during an S1 LOCA, was modeled and qualitative and quantitative comparisons were made to WASH 1400 aand other studies.

  15. Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants

    SciTech Connect (OSTI)

    Simmons, Kevin L.; Ramuhalli, Pradeep; Brenchley, David L.; Coble, Jamie B.; Hashemian, Hash; Konnik, Robert; Ray, Sheila

    2012-09-14T23:59:59.000Z

    The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), NDE instrumentation development, universities, commercial NDE services and cable manufacturers, and Electric Power Research Institute (EPRI). The motivation for the R&D roadmap comes from the need to address the aging management of in-containment cables at nuclear power plants (NPPs).

  16. SciTech Connect: RADTRAD: A simplified model for RADionuclide...

    Office of Scientific and Technical Information (OSTI)

    code developed for the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Reactor Regulation (NRR) to estimate transport and removal of radionuclides and dose at...

  17. ORISE: Applied health physics projects

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

    for government agencies, including the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Reactor Regulation, by supporting the development of technical evaluation reports...

  18. The sequence coding and search system: An approach for constructing and analyzing event sequences at commercial nuclear power plants

    SciTech Connect (OSTI)

    Mays, G.T.

    1989-04-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) has recognized the importance of the collection, assessment, and feedstock of operating experience data from commercial nuclear power plants and has centralized these activities in the Office for Analysis and Evaluation of Operational Data (AEOD). Such data is essential for performing safety and reliability analyses, especially analyses of trends and patterns to identify undesirable changes in plant performance at the earliest opportunity to implement corrective measures to preclude the occurrences of a more serious event. One of NRC's principal tools for collecting and evaluating operating experience data is the Sequence Coding and Search System (SCSS). The SCSS consists of a methodology for structuring event sequences and the requisite computer system to store and search the data. The source information for SCSS is the Licensee Event Report (LER), which is a legally required document. This paper describes the objective SCSS, the information it contains, and the format and approach for constructuring SCSS event sequences. Examples are presented demonstrating the use SCSS to support the analysis of LER data. The SCSS contains over 30,000 LERs describing events from 1980 through the present. Insights gained from working with a complex data system from the initial developmental stage to the point of a mature operating system are highlighted.

  19. NRC Research Program on Plant Aging: Listing and summaries of reports issued through September 1993. Revision 4

    SciTech Connect (OSTI)

    Vora, J.P.

    1993-12-01T23:59:59.000Z

    The US Nuclear Regulatory Commission is conducting the Nuclear Plant Aging Research (NPAR) Program. This is a comprehensive hardware-oriented engineering research program focused on understanding the aging mechanisms of components and systems in nuclear power plants. The NPAR program also focuses on methods for simulating and monitoring the aging-related degradation of these components and systems. In addition, it provides recommendations for effective maintenance to manage aging and for implementation of the research results in the regulatory process. This document contains a listing and index of reports generated in the NPAR Program that were issued through September 1993 and summaries of those reports. Each summary describes the elements of the research covered in the report and outlines the significant results. For the convenience of the user, the reports are indexed by personal author, corporate author, and subject.

  20. Microsoft PowerPoint - 4_JOHN_BALLARD_MARY_MCCDONNELL_NRC DOE...

    National Nuclear Security Administration (NNSA)

    Len Myers Special Requests (301) 903-2180 Len.Myers@nnsa.doe.gov Patricia Smith NRC Reconciliation (301) 903-6860 PatriciaR.Smith@nnsa.doe.gov Reasons to Call NMMSS -...

  1. Audit Report on "The Department's Management of Nuclear Materials Provided to Domestic Licensees"

    SciTech Connect (OSTI)

    None

    2009-02-01T23:59:59.000Z

    The objective if to determine whether the Department of Energy (Department) was adequately managing its nuclear materials provided to domestic licensees. The audit was performed from February 2007 to September 2008 at Department Headquarters in Washington, DC, and Germantown, MD; the Oak Ridge Office and the Oak Ridge National Laboratory in Oak Ridge, TN. In addition, we visited or obtained data from 40 different non-Departmental facilities in various states. To accomplish the audit objective, we: (1) Reviewed Departmental and Nuclear Regulatory Commission (NRC) requirements for the control and accountability of nuclear materials; (2) Analyzed a Nuclear Materials Management and Safeguards System (NMMSS) report with ending inventory balances for Department-owned nuclear materials dated September 30, 2007, to determine the amount and types of nuclear materials located at non-Department domestic facilities; (3) Held discussions with Department and NRC personnel that used NMMSS information to determine their roles and responsibilities related to the control and accountability over nuclear materials; (4) Selected a judgmental sample of 40 non-Department domestic facilities; (5) Met with licensee officials and sent confirmations to determine whether their actual inventories of Department-owned nuclear materials were consistent with inventories reported in the NMMSS; and, (6) Analyzed historical information related to the 2004 NMMSS inventory rebaselining initiative to determine the quantity of Department-owned nuclear materials that were written off from the domestic licensees inventory balances. This performance audit was conducted in accordance with generally accepted Government auditing standards. Those standards require that we plan and perform the audit to obtain sufficient, appropriate evidence to provide a reasonable basis for our findings and conclusions based on our audit objective. We believe that the evidence obtained provides a reasonable basis for our findings and conclusions based on our audit objectives. The audit included tests of controls and compliance with laws and regulations related to managing the Department-owned nuclear materials provided to non-Departmental domestic licensees. Because our review was limited it would not necessarily have disclosed all internal control deficiencies that may have existed at the time of our audit. We examined the establishment of performance measures in accordance with Government Performance and Results Act of 1993, as they related to the audit objective. We found that the Department had established performance measures related to removing or disposing of nuclear materials and radiological sources around the world. We utilized computer generated data during our audit and performed procedures to validate the reliability of the information as necessary to satisfy our audit objective. As noted in the report, we questioned the reliability of the NMMSS data.

  2. Understanding the Challenges in the Transition from Film Radiography in the Nuclear Power Industry

    SciTech Connect (OSTI)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Moran, Traci L.; Nove, Carol A.; Pardini, Allan F.

    2012-09-01T23:59:59.000Z

    Nondestructive examination (NDE) applications in the nuclear power industry using film radiography are shrinking due to the advent of modern digital imaging technologies and advances in alternative inspection methods that do not present an ionizing radiation hazard. Technologies that are used routinely in the medical industry for patient diagnosis are being adapted to industrial NDE applications including the detection and characterization of defects in welds. From the user perspective, non-film inspection techniques provide several advantages over film techniques. It is anticipated that the shift away from the application of film radiography in the nuclear power industry represents an irreversible trend. The U.S. Nuclear Regulatory Commission (NRC) has noted this trend in the U.S. nuclear power industry and will be working to ensure that the effectiveness and reliability of component inspections is not compromised by this transition. Currently, specific concerns are associated with 1) obtaining a fundamental understanding of how inspection effectiveness and reliability may be impacted by this transition and 2) ensuring training standards and qualifications remain compatible with modern industrial radiographic practice. This paper discusses recent trends in industrial radiography and assesses their advantages and disadvantages from the perspective of nuclear power plant component inspections.

  3. Safety research programs sponsored by Office of Nuclear Regulatory Research. Quarterly progress report, October 1-December 31, 1983. Volume 3, No. 4

    SciTech Connect (OSTI)

    Weiss, A J [comp.

    1984-05-01T23:59:59.000Z

    The projects reported are the following: High Temperature Reactor Research, SSC Development, Validation and Application, CRBR Balance of Plant Modeling, Thermal-Hydraulic Reactor Safety Experiments, Development of Plant Analyzer, Code Assessment and Application (Transient and LOCA Analyses), Thermal Reactor Code Development (RAMONA-3B), Calculational Quality Assurance in Support of PTS; Stress Corrosion Cracking of PWR Steam Generator Tubing, Bolting Failure Analysis, Probability Based Load Combinations for Design of Category I Structures, Mechanical Piping Benchmark Problems, Identification of Age-Related Failure Modes; Analysis of Human Error Data for Nuclear Power Plant Safety-Related Events, Human Factors in Nuclear Power Plant Safeguards, Emergency Action Levels, and Protective Action Decision Making.

  4. 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.

  5. Interim Management of Nuclear Materials

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

    by the appropriate agency (DOE, the Department of Transportation, the Nuclear Regulatory Commission, or the International Atomic Energy Agency) for full containment of the...

  6. ADVANCED CERAMIC MATERIALS FOR NEXT-GENERATION NUCLEAR APPLICATIONS

    SciTech Connect (OSTI)

    Marra, J.

    2010-09-29T23:59:59.000Z

    Rising global energy demands coupled with increased environmental concerns point to one solution; they must reduce their dependence on fossil fuels that emit greenhouse gases. As the global community faces the challenge of maintaining sovereign nation security, reducing greenhouse gases, and addressing climate change nuclear power will play a significant and likely growing role. In the US, nuclear energy already provides approximately one-fifth of the electricity used to power factories, offices, homes, and schools with 104 operating nuclear power plants, located at 65 sites in 31 states. Additionally, 19 utilities have applied to the US Nuclear Regulatory Commission (NRC) for construction and operating licenses for 26 new reactors at 17 sites. This planned growth of nuclear power is occurring worldwide and has been termed the 'nuclear renaissance.' As major industrial nations craft their energy future, there are several important factors that must be considered about nuclear energy: (1) it has been proven over the last 40 years to be safe, reliable and affordable (good for Economic Security); (2) its technology and fuel can be domestically produced or obtained from allied nations (good for Energy Security); and (3) it is nearly free of greenhouse gas emissions (good for Environmental Security). Already an important part of worldwide energy security via electricity generation, nuclear energy can also potentially play an important role in industrial processes and supporting the nation's transportation sector. Coal-to-liquid processes, the generation of hydrogen and supporting the growing potential for a greatly increased electric transportation system (i.e. cars and trains) mean that nuclear energy could see dramatic growth in the near future as we seek to meet our growing demand for energy in cleaner, more secure ways. In order to address some of the prominent issues associated with nuclear power generation (i.e., high capital costs, waste management, and proliferation), the worldwide community is working to develop and deploy new nuclear energy systems and advanced fuel cycles. These new nuclear systems address the key challenges and include: (1) extracting the full energy value of the nuclear fuel; (2) creating waste solutions with improved long term safety; (3) minimizing the potential for the misuse of the technology and materials for weapons; (4) continually improving the safety of nuclear energy systems; and (5) keeping the cost of energy affordable.

  7. Physical fitness training reference manual for security force personnel at fuel cycle facilities possessing formula quantities of special nuclear materials

    SciTech Connect (OSTI)

    Arzino, P.A.; Caplan, C.S.; Goold, R.E. (California State Univ., Hayward, CA (United States). Foundation)

    1991-09-01T23:59:59.000Z

    The recommendations contained throughout this NUREG are being provided to the Nuclear Regulatory Commission (NRC) as a reference manual which can be used by licensee management as they develop a program plan for the safe participation of guards, Tactical Response Team members (TRTs), and all other armed response personnel in physical fitness training and in physical performance standards testing. The information provided in this NUREG will help licensees to determine if guards, TRTs, and other armed response personnel can effectively perform their normal and emergency duties without undue hazard to themselves, to fellow employees, to the plant site, and to the general public. The recommendations in this NUREG are similar in part to those contained within the Department of Energy (DOE) Medical and Fitness Implementation Guide which was published in March 1991. The guidelines contained in this NUREG are not requirements, and compliance is not required. 25 refs.

  8. Medical screening reference manual for security force personnel at fuel cycle facilities possessing formula quantities of special nuclear materials

    SciTech Connect (OSTI)

    Arzino, P.A.; Brown, C.H. (California State Univ., Hayward, CA (United States). Foundation)

    1991-09-01T23:59:59.000Z

    The recommendations contained throughout this NUREG were provided to the Nuclear Regulatory Commission (NRC) as medical screening information that could be used by physicians who are evaluating the parameters of the safe participation of guards, Tactical Response Team members (TRTs), and all other armed response personnel in physical fitness training and in physical performance standards testing. The information provided in this NUREG will help licensees to determine if guards, TRTs, and other armed response personnel can effectively perform their normal and emergency duties without undue hazard to themselves, to fellow employees, to the plant site, and to the general public. The medical recommendations in this NUREG are similar in content to the medical standards contained in 10 CFR Part 1046 which, in part, specifies medical standards for the protective force personnel regulated by the Department of Energy. The guidelines contained in this NUREG are not requirements, and compliance is not required. 3 refs.

  9. Plans, Updates, Regulatory Documents

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

    Stewardship Environmental Protection Obeying Environmental Laws Individual Permit Documents Individual Permit: Plans, Updates, Regulatory Documents1335769200000Plans...

  10. ENVIRONMENTAL ASSESSMENT/ REGULATORY IMPACT REVIEW/

    E-Print Network [OSTI]

    ENVIRONMENTAL ASSESSMENT/ REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS Area and Regulatory Amendments for Bering Sea Habitat Conservation May 2008 Lead Agency: National Juneau, AK 99802 (907) 586-7228 Abstract: This Environmental Assessment/Regulatory Impact Review

  11. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 6, Decontamination and decommissioning, accident management, TMI-2

    SciTech Connect (OSTI)

    Weiss, A. J. [comp.

    1988-02-01T23:59:59.000Z

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 6, discusses decontamination and decommissioning, accident management, and the Three Mile Island-2 reactor accident. Thirteen reports have been cataloged separately.

  12. A Review of Sensor Calibration Monitoring for Calibration Interval Extension in Nuclear Power Plants

    SciTech Connect (OSTI)

    Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep; Bond, Leonard J.; Hashemian, Hash; Shumaker, Brent; Cummins, Dara

    2012-08-31T23:59:59.000Z

    Currently in the United States, periodic sensor recalibration is required for all safety-related sensors, typically occurring at every refueling outage, and it has emerged as a critical path item for shortening outage duration in some plants. Online monitoring can be employed to identify those sensors that require calibration, allowing for calibration of only those sensors that need it. International application of calibration monitoring, such as at the Sizewell B plant in United Kingdom, has shown that sensors may operate for eight years, or longer, within calibration tolerances. This issue is expected to also be important as the United States looks to the next generation of reactor designs (such as small modular reactors and advanced concepts), given the anticipated longer refueling cycles, proposed advanced sensors, and digital instrumentation and control systems. The U.S. Nuclear Regulatory Commission (NRC) accepted the general concept of online monitoring for sensor calibration monitoring in 2000, but no U.S. plants have been granted the necessary license amendment to apply it. This report presents a state-of-the-art assessment of online calibration monitoring in the nuclear power industry, including sensors, calibration practice, and online monitoring algorithms. This assessment identifies key research needs and gaps that prohibit integration of the NRC-approved online calibration monitoring system in the U.S. nuclear industry. Several needs are identified, including the quantification of uncertainty in online calibration assessment; accurate determination of calibration acceptance criteria and quantification of the effect of acceptance criteria variability on system performance; and assessment of the feasibility of using virtual sensor estimates to replace identified faulty sensors in order to extend operation to the next convenient maintenance opportunity. Understanding the degradation of sensors and the impact of this degradation on signals is key to developing technical basis to support acceptance criteria and set point decisions, particularly for advanced sensors which do not yet have a cumulative history of operating performance.

  13. NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

    SciTech Connect (OSTI)

    Pinkston, Karen E.; Ridge, A. Christianne; Alexander, George W.; Barr, Cynthia S.; Devaser, Nishka J.; Felsher, Harry D. [U.S. Nuclear Regulatory Commission (United States)] [U.S. Nuclear Regulatory Commission (United States)

    2013-07-01T23:59:59.000Z

    As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in coordination with South Carolina Department of Health and Environmental Control (SCDHEC). DOE has completed or begun additional work related to salt waste disposal to address these factors. NRC staff continues to evaluate information related to the performance of the SDF and has been working with DOE and SCDHEC to resolve NRC staff's technical concerns. (authors)

  14. U.S. Nuclear Power Plant Operating Cost and Experience Summaries

    SciTech Connect (OSTI)

    Reid, RL

    2003-09-18T23:59:59.000Z

    The ''U.S. Nuclear Power Plant Operating Cost and Experience Summaries'' (NUREG/CR-6577, Supp. 2) report has been prepared to provide historical operating cost and experience information on U.S. commercial nuclear power plants during 2000-2001. Costs incurred after initial construction are characterized as annual production costs, which represent fuel and plant operating and maintenance expenses, and capital expenditures related to facility additions/modifications, which are included in the plant capital asset base. As discussed in the report, annual data for these two cost categories were obtained from publicly available reports and must be accepted as having different degrees of accuracy and completeness. Treatment of inconclusive and incomplete data is discussed. As an aid to understanding the fluctuations in the cost histories, operations summaries for each nuclear unit are provided. The intent of these summaries is to identify important operating events; refueling, major maintenance, and other significant outages; operating milestones; and significant licensing or enforcement actions. Information used in the summaries is condensed from operating reports submitted by the licensees, the Nuclear Regulatory Commission (NRC) database for enforcement actions, and outage reports.

  15. Nuclear Regulatory Commission Regulatory and Licensing Matters | Department

    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 theLiability Nuclearof Energy

  16. MARSAME Appendix B B. SOURCES OF BACKGROUND RADIOACTIVITY

    E-Print Network [OSTI]

    : · The Nuclear Regulatory Commission (NRC) provides information concerning background radioactivity in Background as a Residual Radioactivity Criterion for Decommissioning NUREG-1501 (NRC 1994). · The United Nations Scientific

  17. Light Water Reactor Sustainability (LWRS) Program – Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants

    SciTech Connect (OSTI)

    Simmons, K.L.; Ramuhali, P.; Brenchley, D.L.; Coble, J.B.; Hashemian, H.M.; Konnick, R.; Ray, S.

    2012-09-01T23:59:59.000Z

    Executive Summary [partial] The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. A workshop was held to gather subject matter experts to develop the NDE R&D Roadmap for Cables. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, and NDE instrumentation development from the U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), universities, commercial NDE service vendors and cable manufacturers, and the Electric Power Research Institute (EPRI).

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

    E-Print Network [OSTI]

    Yen, W.W.S.

    2010-01-01T23:59:59.000Z

    S. Commercial Nuclear Power Plants. WASH-1400. October 1975.Content of for Nuclear Power Plants. Regulatory Guide 1.101.PLANNING FOR NUCLEAR POWER PLANTS: THE LICENSING PROCESS

  19. DOE, NRC Issue Licensing Roadmap For Next-Generation Nuclear Plant |

    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: Theof"WaveInteractionsMaterialsDevelopEnergyof EnergyDOE, NEPA,

  20. DOE, NRC Issue Licensing Roadmap For Next-Generation Nuclear Plant |

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directed off Energy.gov. Are you0and Transparency, and MoreEnergyof EnergyDepartmentDOE's RoundDepartment

  1. 3 HISTORICAL SITE ASSESSMENT 3.1 Introduction

    E-Print Network [OSTI]

    , some facilities--such as Nuclear Regulatory Commission (NRC) licensees that routinely maintain records decommissioning procedure (see Appendix B). The HSA ! identifies potential, likely, or known sources

  2. ORISE: Health physics services

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

    Nuclear Regulatory Commission (NRC), as well as other federal and state agencies. From radiological facility audits and reviews to dose modeling and technical evaluations, ORISE...

  3. Office of Enterprise Assessments Review of the West Valley Demonostrat...

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

    Liquid Waste Cell NRC Nuclear Regulatory Commission OFI Opportunity for Improvement PM Preventive Maintenance PPE Personal Protective Equipment RBA Radiological Buffer Area RCT...

  4. December 11-12, 2014 - Technical Exchange Meeting - Las Vegas...

    Office of Environmental Management (EM)

    Agency (EPA), U.S. Nuclear Regulatory Commission (NRC), Washington State Department of Ecology, and Commonwealth of Kentucky, conducted the Annual Technical Exchange for the...

  5. Superfund Policy Statements and Guidance Regarding Disposition of Radioactive Waste in Non-NRC Licensed Disposal Facilities - 13407

    SciTech Connect (OSTI)

    Walker, Stuart [U.S. Environmental Protection Agency (United States)] [U.S. Environmental Protection Agency (United States)

    2013-07-01T23:59:59.000Z

    This talk will discuss EPA congressional testimony and follow-up letters, as well as letters to other stakeholders on EPA's perspectives on the disposition of radioactive waste outside of the NRC licensed disposal facility system. This will also look at Superfund's historical practices, and emerging trends in the NRC and agreement states on waste disposition. (author)

  6. NRC Earth Science Decadal Survey-Mission Concept Earth Sciences from the Astronomer's Perspective, a Deep

    E-Print Network [OSTI]

    Christian, Eric

    NRC Earth Science Decadal Survey-Mission Concept Earth Sciences from the Astronomer's Perspective Irina Melnikova #12;1 Earth Sciences from the Astronomer's Perspective 1.0 Mission Concept and Purpose Earth observations from satellites located in deep space offer the exciting opportunity to look

  7. Understanding and Interpreting the NRC's "Data-Based Assessment of Research-Doctorate

    E-Print Network [OSTI]

    Navara, Kristen

    member per year [2000-2006] R Avg. citations per publication [2000-2006] R % Faculty with Grant [2005-06 AY] R % Interdisciplinary Faculty [2005-06 AY] % non-Asian Minority Faculty [2005-06 AY] D % Female Faculty [2005-06 AY] D Awards/Faculty Member [2001-2006] R 10 #12;10/21/2010 6 NRC Methodology

  8. NRC TLD Direct Radiation Monitoring Network. Progress report, October--December 1996

    SciTech Connect (OSTI)

    Struckmeyer, R.

    1997-03-01T23:59:59.000Z

    This report presents the results of the NRC Direct Radiation Monitoring Network for the fourth quarter of 1996. It provides the ambient radiation levels measured in the vicinity of 74 sites throughout the United States. In addition, it describes the equipment used, monitoring station selection criteria, characterization of the dosimeter response, calibration procedures, statistical methods, intercomparison, and quality assurance program. 3 figs., 4 tabs.

  9. NRC TLD Direct Radiation Monitoring Network progress report, October--December 1994. Volume 14, No. 4

    SciTech Connect (OSTI)

    Struckmeyer, R.

    1995-03-01T23:59:59.000Z

    This report presents the results of the NRC Direct Radiation Monitoring Network for the fourth quarter of 1994. It provides the ambient radiation levels measured in the vicinity of 75 sites throughout the United States. In addition, it describes the equipment used, monitoring station selection criteria, characterization of the dosimeter response, calibration procedures, statistical methods, intercomparison, and quality assurance program.

  10. HOW MANY DID YOU SAY? HISTORICAL AND PROJECTED SPENT NUCLEAR FUEL SHIPMENTS IN THE UNITED STATES, 1964 - 2048

    SciTech Connect (OSTI)

    Halstead, Robert J.; Dilger, Fred

    2003-02-27T23:59:59.000Z

    No comprehensive, up-to-date, official database exists for spent nuclear fuel shipments in the United States. The authors review the available data sources, and conclude that the absence of such a database can only be rectified by a major research effort, similar to that carried out by Oak Ridge National Laboratory (ORNL) in the early 1990s. Based on a variety of published references, and unpublished data from the U.S. Nuclear Regulatory Commission (NRC), the authors estimate cumulative U.S. shipments of commercial spent fuel for the period 1964-2001. The cumulative estimates include quantity shipped, number of cask-shipments, and shipment-miles, by truck and by rail. The authors review previous estimates of future spent fuel shipments, including contractor reports prepared for the U.S. Department of Energy (DOE), NRC, and the State of Nevada. The DOE Final Environmental Impact Statement (FEIS) for Yucca Mountain includes projections of spent nuclear fuel and high-level radioactive was te shipments for two inventory disposal scenarios (24 years and 38 years) and two national transportation modal scenarios (''mostly legal-weight truck'' and ''mostly rail''). Commercial spent fuel would compromise about 90 percent of the wastes shipped to the repository. The authors estimate potential shipments to Yucca Mountain over 38 years (2010-2048) for the DOE ''mostly legal-weight truck'' and ''mostly rail'' scenarios, and for an alternative modal mix scenario based on current shipping capabilities of the 72 commercial reactor sites. The cumulative estimates of future spent fuel shipments include quantity shipped, number of cask-shipments, and shipment-miles, by legal-weight truck, heavy-haul truck, rail and barge.

  11. ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS

    E-Print Network [OSTI]

    ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS.0 NEPA REQUIREMENTS: ENVIRONMENTAL IMPACTS OF THE ALTERNATIVES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.1 Environmental Impacts of the Alternatives

  12. Phenylpropanoid related regulatory protein-regulatory region associations

    DOE Patents [OSTI]

    Apuya, Nestor (Culver City, CA); Bobzin, Steven Craig (Malibu, CA); Park, Joon-Hyun (Oak Park, CA); Doukhanina, Elena (Newbury Park, CA)

    2012-01-03T23:59:59.000Z

    Materials and methods for identifying lignin regulatory region-regulatory protein associations are disclosed. Materials and methods for modulating lignin accumulation are also disclosed.

  13. Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

    SciTech Connect (OSTI)

    OHara J. M.; Higgins, J.; DAgostino, A.

    2012-01-17T23:59:59.000Z

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

  14. Compression and immersion tests and leaching of radionuclides, stable metals, and chelating agents from cement-solidified decontamination waste collected from nuclear power stations

    SciTech Connect (OSTI)

    Akers, D.W.; Kraft, N.C.; Mandler, J.W. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-06-01T23:59:59.000Z

    A study was performed for the Nuclear Regulatory Commission (NRC) to evaluate structural stability and leachability of radionuclides, stable metals, and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from seven commercial boiling water reactors and one pressurized water reactor. The decontamination methods used at the reactors were the Can-Decon, AP/Citrox, Dow NS-1, and LOMI processes. Samples of untreated resin waste and solidified waste forms were subjected to immersion and compressive strength testing. Some waste-form samples were leach-tested using simulated groundwaters and simulated seawater for comparison with the deionized water tests that are normally performed to assess waste-form leachability. This report presents the results of these tests and assesses the effects of the various decontamination methods, waste form formulations, leachant chemical compositions, and pH of the leachant on the structural stability and leachability of the waste forms. Results indicate that releases from intact and degraded waste forms are similar and that the behavior of some radionuclides such as {sup 55}Fe, {sup 60}Co, and {sup 99}Tc were similar. In addition, the leachability indexes are greater than 6.0, which meets the requirement in the NRC`s ``Technical Position on Waste Form,`` Revision 1.

  15. Radioactive Semivolatiles in Nuclear Fuel Reprocessing

    SciTech Connect (OSTI)

    R. T. Jubin; D. M. Strachan; G. Ilas; B. B. Spencer; N. R. Soelberg

    2014-09-01T23:59:59.000Z

    In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14C, 3H, 129I, and 85Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is then discussed with respect to what is known in the literature about their behavior in a reprocessing facility. The context for the evaluation in this document is a UO2-based fuel processed through an aqueous-based reprocessing system with a TBP-based solvent extraction chemistry. None of these elements form sufficiently volatile compounds in the context of the reprocessing facility to be of regulatory concern.

  16. Central and Eastern United States (CEUS) Seismic Source Characterization (SSC) for Nuclear Facilities Project

    SciTech Connect (OSTI)

    Kevin J. Coppersmith; Lawrence A. Salomone; Chris W. Fuller; Laura L. Glaser; Kathryn L. Hanson; Ross D. Hartleb; William R. Lettis; Scott C. Lindvall; Stephen M. McDuffie; Robin K. McGuire; Gerry L. Stirewalt; Gabriel R. Toro; Robert R. Youngs; David L. Slayter; Serkan B. Bozkurt; Randolph J. Cumbest; Valentina Montaldo Falero; Roseanne C. Perman' Allison M. Shumway; Frank H. Syms; Martitia (Tish) P. Tuttle [Tish

    2012-01-31T23:59:59.000Z

    This report describes a new seismic source characterization (SSC) model for the Central and Eastern United States (CEUS). It will replace the Seismic Hazard Methodology for the Central and Eastern United States, EPRI Report NP-4726 (July 1986) and the Seismic Hazard Characterization of 69 Nuclear Plant Sites East of the Rocky Mountains, Lawrence Livermore National Laboratory Model, (Bernreuter et al., 1989). The objective of the CEUS SSC Project is to develop a new seismic source model for the CEUS using a Senior Seismic Hazard Analysis Committee (SSHAC) Level 3 assessment process. The goal of the SSHAC process is to represent the center, body, and range of technically defensible interpretations of the available data, models, and methods. Input to a probabilistic seismic hazard analysis (PSHA) consists of both seismic source characterization and ground motion characterization. These two components are used to calculate probabilistic hazard results (or seismic hazard curves) at a particular site. This report provides a new seismic source model. Results and Findings The product of this report is a regional CEUS SSC model. This model includes consideration of an updated database, full assessment and incorporation of uncertainties, and the range of diverse technical interpretations from the larger technical community. The SSC model will be widely applicable to the entire CEUS, so this project uses a ground motion model that includes generic variations to allow for a range of representative site conditions (deep soil, shallow soil, hard rock). Hazard and sensitivity calculations were conducted at seven test sites representative of different CEUS hazard environments. Challenges and Objectives The regional CEUS SSC model will be of value to readers who are involved in PSHA work, and who wish to use an updated SSC model. This model is based on a comprehensive and traceable process, in accordance with SSHAC guidelines in NUREG/CR-6372, Recommendations for Probabilistic Seismic Hazard Analysis: Guidance on Uncertainty and Use of Experts. The model will be used to assess the present-day composite distribution for seismic sources along with their characterization in the CEUS and uncertainty. In addition, this model is in a form suitable for use in PSHA evaluations for regulatory activities, such as Early Site Permit (ESPs) and Combined Operating License Applications (COLAs). Applications, Values, and Use Development of a regional CEUS seismic source model will provide value to those who (1) have submitted an ESP or COLA for Nuclear Regulatory Commission (NRC) review before 2011; (2) will submit an ESP or COLA for NRC review after 2011; (3) must respond to safety issues resulting from NRC Generic Issue 199 (GI-199) for existing plants and (4) will prepare PSHAs to meet design and periodic review requirements for current and future nuclear facilities. This work replaces a previous study performed approximately 25 years ago. Since that study was completed, substantial work has been done to improve the understanding of seismic sources and their characterization in the CEUS. Thus, a new regional SSC model provides a consistent, stable basis for computing PSHA for a future time span. Use of a new SSC model reduces the risk of delays in new plant licensing due to more conservative interpretations in the existing and future literature. Perspective The purpose of this study, jointly sponsored by EPRI, the U.S. Department of Energy (DOE), and the NRC was to develop a new CEUS SSC model. The team assembled to accomplish this purpose was composed of distinguished subject matter experts from industry, government, and academia. The resulting model is unique, and because this project has solicited input from the present-day larger technical community, it is not likely that there will be a need for significant revision for a number of years. See also Sponsors Perspective for more details. The goal of this project was to implement the CEUS SSC work plan for developing a regional CEUS SSC model. The work plan, formulated by the project manager and a

  17. ORNL/TM-2005/553 The Effect of Elevated Temperature on

    E-Print Network [OSTI]

    Pennycook, Steve

    for the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Under Interagency Agreement.4) Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Under Interagency Agreement No. 1886-N674-1Y NRC FIN No. Y6741 #12;This report was prepared as an account of work sponsored

  18. Directory of Certificates of Compliance for radioactive materials packages: Report of NRC approved packages. Volume 1, Revision 18

    SciTech Connect (OSTI)

    NONE

    1995-10-01T23:59:59.000Z

    The purpose of this directory is to make available a convenient source of information on packagings which have been approved by the US Nuclear Regulatory Commission. To assist in identifying packaging, an index by Model Number and corresponding Certificate of Compliance Number is included at the front of Volumes 1 and 2. An alphabetical listing by user name is included in the back of Volume 3 of approved QA programs. The reports include a listing of all users of each package design and approved QA programs prior to the publication date.

  19. Title list of documents made publicly available, June 1-30, 1997

    SciTech Connect (OSTI)

    NONE

    1997-08-01T23:59:59.000Z

    This document is a monthly publication describing information received and published by the U.S. Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and published by NRC pertinent to its role as a regulatory agency. As used here, {open_quotes}docketed{close_quotes} does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records.

  20. REGULATORY STATUS: AOC

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

    184 UNIT NAME: Concrete Rubble pile 129) REGULATORY STATUS: AOC LOCATION: Outside plant security south of C-611 Water Treatment Plant. Estimated to be less than 5 cubic feet....

  1. Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges

    SciTech Connect (OSTI)

    B. McLeod

    2002-02-28T23:59:59.000Z

    This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M&O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M&O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report summarizes the results of the 2002 Reference SNF Discharge Projection.

  2. Title list of documents made publicly available: February 1--28, 1995. Volume 17, Number 2

    SciTech Connect (OSTI)

    NONE

    1995-04-01T23:59:59.000Z

    This monthly publication contains descriptions of the information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. As used here, docketed refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index. NRC documents that are publicly available may be examined without charge at the NRC Public Document Room (PDR).

  3. U.S. NRC CONFIRMATORY LEVEL 1 PRA SUCCESS CRITERIA ACTIVITIES

    SciTech Connect (OSTI)

    Donald Helton; Hossein Esmaili; Robert Buell

    2011-03-01T23:59:59.000Z

    The U.S. Nuclear Regulatory Commission’s standardized plant analysis risk (SPAR) models are used to support a number of risk-informed initiatives. The fidelity and realism of these models are ensured through a number of processes including cross-comparison with industry models, review and use by a wide range of technical experts, and confirmatory analysis. This paper will describe a key activity in the latter arena. Specifically, this paper will describe MELCOR analyses performed to augment the technical basis for confirming or modifying specific success criteria of interest. The analyses that will be summarized provide the basis for confirming or changing success criteria in a specific 3-loop pressurized-water reactor and a Mark-I boiling-water reactor. Initiators that have been analyzed include loss-of-coolant accidents, loss of main feedwater, spontaneous steam generator tube rupture, inadvertent opening of a relief valve at power, and station blackout. For each initiator, specific aspects of the accident evolution are investigated via a targeted set of calculations (3 to 22 distinct accident analyses per initiator). Further evaluation is ongoing to extend the analyses’ conclusions to similar plants (where appropriate), with consideration of design and modeling differences on a scenario-by-scenario basis. This paper will also describe future plans.

  4. A UNITED STATES NUCLEAR REGULATORY COMMISSION

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

    Selection and Evaluation 5-5 5.2.2 Containment Performance 5-9 5.2.3 Control Room and Remote Shutdown Area 5-11 Design 5.2.4 Emergency Planning 5-13 5.2.5 Operator Staffing and...

  5. Nuclear Safety Regulatory Framework | Department of Energy

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

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

  6. NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555

    Office of Legacy Management (LM)

    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 group currentBradleyTableSelling7 August 2008

  7. Nuclear Regulatory Commission | 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 Strategic2 OPAM615_CostNSAR - T en Y earEnergy Research andSite Tritium

  8. UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I

    Office of Legacy Management (LM)

    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 groupTuba City, Arizona, DisposalFourthN V4100 DOE/EA-1452D E P A R T M E

  9. Nuclear Regulatory Commission | 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 of

  10. 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.

  11. COMPARISON OF RESULTS FOR QUARTER 4 SURFACE WATER SPLIT SAMPLES COLLECTED AT THE NUCLEAR FUELS SERVICES SITE, ERWIN, TN

    SciTech Connect (OSTI)

    none,

    2013-08-15T23:59:59.000Z

    Oak Ridge Associated Universities (ORAU), under the Oak Ridge Institute for Science and Education (ORISE) contract, collected split surface water samples with Nuclear Fuel Services (NFS) representatives on June 12, 2013. Representatives from the U.S. Nuclear Regulatory Commission (NRC) and the Tennessee Department of Environment and Conservation were also in attendance. Samples were collected at four surface water stations, as required in the approved Request for Technical Assistance number 11-018. These stations included Nolichucky River upstream (NRU), Nolichucky River downstream (NRD), Martin Creek upstream (MCU), and Martin Creek downstream (MCD). Both ORAU and NFS performed gross alpha and gross beta analyses, and Table 1 presents the comparison of results using the duplicate error ratio (DER), also known as the normalized absolute difference. A DER ≤ 3 indicates at a 99% confidence interval that split sample results do not differ significantly when compared to their respective one standard deviation (sigma) uncertainty (ANSI N42.22). The NFS split sample report specifies 95% confidence level of reported uncertainties (NFS 2013). Therefore, standard two sigma reporting values were divided by 1.96. In conclusion, most DER values were less than 3 and results are consistent with low (e.g., background) concentrations. The gross beta result for sample 5198W0014 was the exception. The ORAU gross beta result of 6.30 ? 0.65 pCi/L from location NRD is well above NFS?s non-detected result of 1.56 ? 0.59 pCi/L. NFS?s data package includes no detected result for any radionuclide at location NRD. At NRC?s request, ORAU performed gamma spectroscopic analysis of sample 5198W0014 to identify analytes contributing to the relatively elevated gross beta results. This analysis identified detected amounts of naturally-occurring constituents, most notably Ac-228 from the thorium decay series, and does not suggest the presence of site-related contamination.

  12. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1994. Twenty-seventh annual report

    SciTech Connect (OSTI)

    Thomas, M.L.; Hagemeyer, D. [Science Applications International Corporation, Oak Ridge, TN (United States)

    1996-01-01T23:59:59.000Z

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). Annual reports for 1994 were received from a total of 303 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 303 licensees indicated that 152,028 individuals were monitored, 79,780 of whom received a measurable dose. The collective dose incurred by these individuals was 24,740 person-cSv (person-rem){sup 2} which represents a 15% decrease from the 1993 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.31 cSv (rem) for 1994. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. In 1994, the annual collective dose per reactor for light water reactor licensees (LWRs) was 198 person-cSv (person-rem). This represents a 18% decrease from the 1993 value of 242 person-cSv (person-rem). The annual collective dose per reactor for boiling water reactors (BWRs) was 327 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 131 person-cSv (person-rem). Analyses of transient worker data indicate that 18,178 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1994, the average measurable dose calculated from reported data was 0.28 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.31 cSv (rem).

  13. 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.

  14. Preparing Non-nuclear Engineers for the Nuclear Field

    E-Print Network [OSTI]

    Ervin, Elizabeth K.

    with nuclear industrial experience. The author intends to publish lessons learned subsequent to the course. Minimally meeting this educational need, a nuclear-related introductory course complies with common, which has been covered at the University of Mississippi by a Nuclear Regulatory Commission Educational

  15. Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.

    SciTech Connect (OSTI)

    Durbin, Samuel G.; Morrow, Charles W.

    2013-01-01T23:59:59.000Z

    The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level - 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

  16. NARAC Modeling During the Response to the Fukushima Dai-ichi Nuclear Power Plant Emergency

    SciTech Connect (OSTI)

    Sugiyama, G; Nasstrom, J S; Probanz, B; Foster, K T; Simpson, M; Vogt, P; Aluzzi, F; Dillon, M; Homann, S

    2012-02-14T23:59:59.000Z

    This paper summarizes the activities of the National Atmospheric Release Advisory Center (NARAC) during the Fukushima Dai-ichi nuclear power plant crisis. NARAC provided a wide range of products and analyses as part of its support including: (1) Daily Japanese weather forecasts and hypothetical release (generic source term) dispersion predictions to provide situational awareness and inform planning for U.S. measurement data collection and field operations; (2) Estimates of potential dose in Japan for hypothetical scenarios developed by the Nuclear Regulatory Commission (NRC) to inform federal government considerations of possible actions that might be needed to protect U.S. citizens in Japan; (3) Estimates of possible plume arrival times and dose for U.S. locations; and (4) Plume model refinement and source estimation based on meteorological analyses and available field data. The Department of Energy/National Nuclear Security Administration (DOE/NNSA) deployed personnel to Japan and stood up 'home team' assets across the DOE complex to aid in assessing the consequences of the releases from the Fukushima Dai-ichi Nuclear Power Plant. The DOE Nuclear Incident Team (NIT) coordinated response activities, while DOE personnel provided predictive modeling, air and ground monitoring, sample collection, laboratory analysis, and data assessment and interpretation. DOE deployed the Aerial Measuring System (AMS), Radiological Assistance Program (RAP) personnel, and the Consequence Management Response Team (CMRT) to Japan. DOE/NNSA home team assets included the Consequence Management Home Team (CMHT); National Atmospheric Release Advisory Center (NARAC); Radiation Emergency Assistance Center/Training Site (REAC/TS); and Radiological Triage. NARAC was activated by the DOE/NNSA on March 11, shortly after the Tohoku earthquake and tsunami occurred. The center remained on active operations through late May when DOE ended its deployment to Japan. Over 32 NARAC staff members, supplemented by other LLNL scientists, invested over 5000 person-hours of time and generated over 300 analyses and predictions.

  17. Best Practices for Finite Element Analysis of Spent Nuclear Fuel Transfer, Storage, and Transportation Systems

    SciTech Connect (OSTI)

    Bajwa, Christopher S.; Piotter, Jason; Cuta, Judith M.; Adkins, Harold E.; Klymyshyn, Nicholas A.; Fort, James A.; Suffield, Sarah R.

    2010-08-11T23:59:59.000Z

    Storage casks and transportation packages for spent nuclear fuel (SNF) are designed to confine SNF in sealed canisters or casks, provide structural integrity during accidents, and remove decay through a storage or transportation overpack. The transfer, storage, and transportation of SNF in dry storage casks and transport packages is regulated under 10 CFR Part 72 and 10 CFR Part 71, respectively. Finite Element Analysis (FEA) is used with increasing frequency in Safety Analysis Reports and other regulatory technical evaluations related to SNF casks and packages and their associated systems. Advances in computing power have made increasingly sophisticated FEA models more feasible, and as a result, the need for careful review of such models has also increased. This paper identifies best practice recommendations that stem from recent NRC review experience. The scope covers issues common to all commercially available FEA software, and the recommendations are applicable to any FEA software package. Three specific topics are addressed: general FEA practices, issues specific to thermal analyses, and issues specific to structural analyses. General FEA practices covers appropriate documentation of the model and results, which is important for an efficient review process. The thermal analysis best practices are related to cask analysis for steady state conditions and transient scenarios. The structural analysis best practices are related to the analysis of casks and associated payload during standard handling and drop scenarios. The best practices described in this paper are intended to identify FEA modeling issues and provide insights that can help minimize associated uncertainties and errors, in order to facilitate the NRC licensing review process.

  18. 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).

  19. Title list of documents made publicly available July 1-31, 1997

    SciTech Connect (OSTI)

    NONE

    1997-09-01T23:59:59.000Z

    This monthly publication describes the information received and published by the U.S. Nuclear Regulatory Commission (US NRC). This includes information on docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and non-docketed material published by the US NRC pertinent to it`s role as a regulatory agency.

  20. PAVAN: an atmospheric-dispersion program for evaluating design-basis accidental releases of radioactive materials from nuclear power stations

    SciTech Connect (OSTI)

    Bander, T.J.

    1982-11-01T23:59:59.000Z

    This report provides a user's guide for the NRC computer program, PAVAN, which is a program used by the US Nuclear Regulatory Commission to estimate downwind ground-level air concentrations for potential accidental releases of radioactive material from nuclear facilities. Such an assessment is required by 10 CFR Part 100 and 10 CFR Part 50. The program implements the guidance provided in Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. Using joint frequency distributions of wind direction and wind speed by atmospheric stability, the program provides relative air concentration (X/Q) values as functions of direction for various time periods at the exclusion area boundary (EAB) and the outer boundary of the low population zone (LPZ). Calculations of X/Q values can be made for assumed ground-level releases (e.g., through building penetrations and vents) or elevated releases from free-standing stacks. Various options may be selected by the user. They can account for variation in the location of release points, additional plume dispersion due to building wakes, plume meander under low wind speed conditions, and adjustments to consider non-straight trajectories. It computes an effective plume height using the physical release height which can be reduced by inputted terrain features. It cannot handle multiple emission sources. A description of the main program and all subroutines is provided. Also included as appendices are a complete listing of the program and two test cases with the required data inputs and the resulting program outputs.

  1. Standardized DOE Spent Nuclear Fuel Canister and Transportation System for Shipping to the National Repository

    SciTech Connect (OSTI)

    Pincock, David Lynn; Morton, Dana Keith; Lengyel, Arpad Leslie

    2001-02-01T23:59:59.000Z

    The U.S.Department of Energy’s (DOE) National Spent Nuclear Fuel Program (NSNFP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), has been chartered with the responsibility for developing spent nuclear fuel (SNF) standardized canisters and a transportation cask system for shipping DOE SNF to the national repository. The mandate for this development is outlined in the Memorandum of Agreement for Acceptance of Department of Energy Spent Nuclear Fuel and High-Level Radioactive Waste that states, “EM shall design and fabricate … DOE SNF canisters for shipment to RW.” (1) It also states, “EM shall be responsible for the design, NRC certification, and fabrication of the transportation cask system for DOE SNF canisters or bare DOE SNF in accordance with 10 CFR Part 71.” (2) In fulfillment of these requirements, the NSNFP has developed four SNF standardized canister configurations and has conceptually designed a versatile transportation cask system for shipping the canisters to the national repository.1 The standardized canister sizes were derived from the national repository waste package design for co-disposal of SNF with high-level waste (HLW). One SNF canister can be placed in the center of the waste package or one can be placed in one of five radial positions, replacing a HLW canister. The internal cavity of the transportation cask was derived using the same logic, matching the size of the internal cavity of the waste package. The size of the internal cavity for the transportation cask allows the shipment of multiple canister configurations with the application of a removable basket design. The standardized canisters have been designed to be loaded with DOE SNF, placed into interim storage, shipped to the national repository, and placed in a waste package without having to be reopened. Significant testing has been completed that clearly demonstrates that the standardized canisters can safely achieve their intended design goals. The transportation cask system will include all of the standard design features, with the addition of dual containment for the shipment of failed fuel. The transportation cask system will also meet the rigorous licensing requirements of the Nuclear Regulatory Commission (NRC) to ensure that the design and the methods of fabrication employed will result in a shipping cask that will safely contain the radioactive materials under all credible accident scenarios. The standardization of the SNF canisters and the versatile design of the transportation cask system will eliminate a proliferation of designs and simplify the operations at the user sites and the national repository.

  2. Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing. Volume 2, Session 3A--Session 4B

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    The 1994 Symposium on Valve and Pump Testing, jointly sponsored by the Board of Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. Individual papers have been cataloged separately.

  3. Proceedings of the Third NRC/ASME Symposium on Valve and Pump Testing. Session 1A--Session 2C: Volume 1

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    The 1994 Symposium on Valve and Pump Testing, jointly sponsored by the Board of Nuclear Codes and Standards of the American Society of Mechanical Engineers and by the Nuclear Regulatory Commission, provides a forum for the discussion of current programs and methods for inservice testing and motor-operated valve testing at nuclear power plants. The symposium also provides an opportunity to discuss the need to improve that testing in order to help ensure the reliable performance of pumps and valves. The participation of industry representatives, regulators, and consultants results in the discussion of a broad spectrum of ideas and perspectives regarding the improvement of inservice testing of pumps and valves at nuclear power plants. This document, Volume 1, covers sessions 1A through session 2C. The individual papers have been cataloged separately.

  4. Mr. John Kinneman, Chief Nuclear Materfals Branch Nuclear Regulatory Commission

    Office of Legacy Management (LM)

    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 group currentBradleyTableSelling7 August 2008 Office7-TACi+J-UN 2 0a50b111989 Mr.

  5. Interstate Technology & Regulatory Council (ITRC) Remediation...

    Office of Environmental Management (EM)

    Technology & Regulatory Council (ITRC) Remediation Management of Complex Sites: Case Studies and Guidance Interstate Technology & Regulatory Council (ITRC) Remediation...

  6. ORISE: Message Testing for a Nuclear Detonation | How ORISE is...

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

    Security, Federal Emergency Management Agency, the Environmental Protection Agency, the Nuclear Regulatory Commission, Department of Energy, Department of Defense, Department of...

  7. Four NNSA teams receive Secretarial Honor Awards | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Team, NNSA Production Office (NPO) Management & Operating (M&O) Contract Placement Team, Nuclear Regulatory Commission Review Support Team, and the USUK Mutual Defense Agreement...

  8. Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 7

    SciTech Connect (OSTI)

    NONE

    1988-01-01T23:59:59.000Z

    The Yucca Mountain site in Neavada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended and approved for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in accordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site;to describe the conceptual designs for the repository and the waste package;and to present the plans for obtaining hte geologic information necessary to demonstrate the suitability of the site for a repository, to design the repository and the waste package, to prepare and environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and developing a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next;it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing principles, and organization of this site characterization plan are outlined, and compliance with applicable regulations is discussed.

  9. Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 4

    SciTech Connect (OSTI)

    NONE

    1988-01-01T23:59:59.000Z

    The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended and approved by the President for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in accordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site; to describe the conceptual designs for the repository and the waste package; and to present the plans for obtaining the geologic information necessary to demonstate the suitability of the site for a repository, to desin the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and developing a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next; it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing principles, and organization of this site characterization plan are outlined, and compliance with applicable regulations is discussed.

  10. Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 2

    SciTech Connect (OSTI)

    NONE

    1988-01-01T23:59:59.000Z

    The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in accordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site; to describe the conceptual designs for the repository and the waste package and to present the plans for obtaining the geologic information necessary to demonstrate the suitability of the site for a repository, to design the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. Chapter 3 summarizes present knowledge of the regional and site hydrologic systems. The purpose of the information presented is to (1) describe the hydrology based on available literature and preliminary site-exploration activities that have been or are being performed and (2) provide information to be used to develop the hydrologic aspects of the planned site characterization program. Chapter 4 contains geochemical information about the Yucca Mountain site. The chapter references plan for continued collection of geochemical data as a part of the site characterization program. Chapter 4 describes and evaluates data on the existing climate and site meterology, and outlines the suggested procedures to be used in developing and validating methods to predict future climatic variation. 534 refs., 100 figs., 72 tabs.

  11. Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 1

    SciTech Connect (OSTI)

    NONE

    1988-01-01T23:59:59.000Z

    The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in acordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site;to describe the conceptual designs for the repository and the waste package and to present the plans for obtaining the geologic information necessary to demonstrate the suitability of the site for a repository, to design the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and eveloping a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next;it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing prinicples, and organization of this site characterization plan are outlined, and compliance with applicable regulations is discussed. 880 refs., 130 figs., 25 tabs.

  12. Standard guide for evaluation of materials used in extended service of interim spent nuclear fuel dry storage systems

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01T23:59:59.000Z

    1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of t...

  13. India's Nuclear Energy Program : prospects The talk will begin with a brief introduction to nuclear fission

    E-Print Network [OSTI]

    Shyamasundar, R.K.

    India's Nuclear Energy Program : prospects The talk will begin with a brief introduction to nuclear posed by reactors, the accident liability laws and regulatory structure governing nuclear energy, Wednesday, Oct 29th 4:00 PM (Tea/Coffee at Seminar Hall, TCIS Colloquium India's Nuclear Energy Program

  14. ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS

    E-Print Network [OSTI]

    ENVIRONMENTAL ASSESSMENT/REGULATORY IMPACT REVIEW/ FINAL REGULATORY FLEXIBILITY ANALYSIS Amendment and Need The purpose of the non-AFA crab sideboard limits was to prevent vessels with crab QS from paper of all GOA sideboards for non-American Fisheries Act (AFA) crab vessels. In April 2007

  15. Nuclear Material Control and Accountability

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

    2011-06-27T23:59:59.000Z

    This Order establishes performance objectives, metrics, and requirements for developing, implementing, and maintaining a nuclear material control and accountability program within DOE/NNSA and for DOE-owned materials at other facilities that are exempt from licensing by the Nuclear Regulatory Commission. Cancels DOE M 470.4-6. Admin Chg 1, 8-3-11.

  16. NRC TLD Direct Radiation Monitoring Network. Volume 15, No. 4: Quarterly progress report, October--December 1995

    SciTech Connect (OSTI)

    Struckmeyer, R.

    1996-03-01T23:59:59.000Z

    This report presents the results of the NRC Direct Radiation Monitoring Network for the fourth quarter of 1995. It provides the ambient radiation levels measured in the vicinity of 75 sites throughout the United States. In addition, it describes the equipment used, monitoring station selection criteria, characterization of the dosimeter response, calibration procedures, statistical methods, intercomparison, and quality assurance program.

  17. Paper ID #5951 Evaluating Online Tutorials for Data Structures and Algorithms Courses

    E-Print Network [OSTI]

    Shaffer, Clifford A.

    for online/Distance delivery at the ME Department at VT. In 2010, with an education grant from Nuclear Regulatory Com- mission (NRC) she completed the online design of the Graduate nuclear engineering certificate program. In 2011, the new education grant from NRC, allowed initiating the design of two new nuclear

  18. 25508 Federal Register / Vol. 65, No. 85 / Tuesday, May 2, 2000 / Notices Week of May 29--Tentative

    E-Print Network [OSTI]

    ­11012 Filed 4­28­00; 2:18 pm] BILLING CODE 7590­01­M NUCLEAR REGULATORY COMMISSION Termination of Agreement, the section 274i agreement between the State of Louisiana and the U.S. Nuclear Regulatory Commission (NRC Between the U.S. Nuclear Regulatory Commission and the State of Louisiana Pursuant to Section 274i

  19. Development, analysis, and evaluation of a commercial software framework for the study of Extremely Low Probability of Rupture (xLPR) events at nuclear power plants.

    SciTech Connect (OSTI)

    Kalinich, Donald A.; Helton, Jon Craig; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-12-01T23:59:59.000Z

    Sandia National Laboratories (SNL) participated in a Pilot Study to examine the process and requirements to create a software system to assess the extremely low probability of pipe rupture (xLPR) in nuclear power plants. This project was tasked to develop a prototype xLPR model leveraging existing fracture mechanics models and codes coupled with a commercial software framework to determine the framework, model, and architecture requirements appropriate for building a modular-based code. The xLPR pilot study was conducted to demonstrate the feasibility of the proposed developmental process and framework for a probabilistic code to address degradation mechanisms in piping system safety assessments. The pilot study includes a demonstration problem to assess the probability of rupture of DM pressurizer surge nozzle welds degraded by primary water stress-corrosion cracking (PWSCC). The pilot study was designed to define and develop the framework and model; then construct a prototype software system based on the proposed model. The second phase of the project will be a longer term program and code development effort focusing on the generic, primary piping integrity issues (xLPR code). The results and recommendations presented in this report will be used to help the U.S. Nuclear Regulatory Commission (NRC) define the requirements for the longer term program.

  20. Handbook for quick cost estimates. A method for developing quick approximate estimates of costs for generic actions for nuclear power plants

    SciTech Connect (OSTI)

    Ball, J.R.

    1986-04-01T23:59:59.000Z

    This document is a supplement to a ''Handbook for Cost Estimating'' (NUREG/CR-3971) and provides specific guidance for developing ''quick'' approximate estimates of the cost of implementing generic regulatory requirements for nuclear power plants. A method is presented for relating the known construction costs for new nuclear power plants (as contained in the Energy Economic Data Base) to the cost of performing similar work, on a back-fit basis, at existing plants. Cost factors are presented to account for variations in such important cost areas as construction labor productivity, engineering and quality assurance, replacement energy, reworking of existing features, and regional variations in the cost of materials and labor. Other cost categories addressed in this handbook include those for changes in plant operating personnel and plant documents, licensee costs, NRC costs, and costs for other government agencies. Data sheets, worksheets, and appropriate cost algorithms are included to guide the user through preparation of rough estimates. A sample estimate is prepared using the method and the estimating tools provided.

  1. Regulatory Streamlining and Improvement

    SciTech Connect (OSTI)

    Mark A. Carl

    2006-07-11T23:59:59.000Z

    The Interstate Oil and Gas Compact Commission (IOGCC) engaged in numerous projects outlined under the scope of work discussed in the United States Department of Energy (DOE) grant number DE-FC26-04NT15456 awarded to the IOGCC. Numerous projects were completed that were extremely valuable to state oil and gas agencies as a result of work performed utilizing resources provided by the grant. There are numerous areas in which state agencies still need assistance. This additional assistance will need to be addressed under future scopes of work submitted annually to DOE's Project Officer for this grant. This report discusses the progress of the projects outlined under the grant scope of work for the 2005-2006 areas of interest, which are as follows: Area of Interest No. 1--Regulatory Streamlining and Improvement: This area of interest continues to support IOGCC's regulatory streamlining efforts that include the identification and elimination of unnecessary duplications of efforts between and among state and federal programs dealing with exploration and production on public lands. Area of Interest No. 2--Technology: This area of interest seeks to improve efficiency in states through the identification of technologies that can reduce costs. Area of Interest No. 3--Training and Education: This area of interest is vital to upgrading the skills of regulators and industry alike. Within the National Energy Policy, there are many appropriate training and education opportunities. Education was strongly endorsed by the President's National Energy Policy Development group. Acting through the governors offices, states are very effective conduits for the dissemination of energy education information. While the IOGCC favors the development of a comprehensive, long-term energy education plan, states are also supportive of immediate action on important concerns, such as energy prices, availability and conservation. Area of Interest No. 4--Resource Assessment and Development: This area of interest relates directly to helping maximize production of domestic oil and natural gas resources, including areas that are under explored or have not been adequately defined.

  2. UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD

    E-Print Network [OSTI]

    UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington are pleased to transmit a technical report prepared by the Nuclear Waste Technical Review Board (Board. Based on its review of data gathered by the DOE and the Center for Nuclear Waste Regulatory Analyses

  3. The Past, Present, and Future of Nuclear

    E-Print Network [OSTI]

    Wechsler, Risa H.

    Commission Regulation and promotional pieces split into the Department of Energy and the Nuclear Regulatory basis WASH-1400, "Reactor Safety Study ­ An Assessment of Accident Risks in U.S. Commercial Nuclear in the vicinity of a nuclear power plant of prompt fatalities that might result from reactor accidents should

  4. High Level Requirements for the Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

    SciTech Connect (OSTI)

    Rich Johnson; Hyung Lee; Kimberlyn C. Mousseau

    2011-09-01T23:59:59.000Z

    The US Department of Energy, Office of Nuclear Energy (DOE-NE), has been tasked with the important mission of ensuring that nuclear energy remains a compelling and viable energy source in the U.S. The motivations behind this mission include cost-effectively meeting the expected increases in the power needs of the country, reducing carbon emissions and reducing dependence on foreign energy sources. In the near term, to ensure that nuclear power remains a key element of U.S. energy strategy and portfolio, the DOE-NE will be working with the nuclear industry to support safe and efficient operations of existing nuclear power plants. In the long term, to meet the increasing energy needs of the U.S., the DOE-NE will be investing in research and development (R&D) and working in concert with the nuclear industry to build and deploy new, safer and more efficient nuclear power plants. The safe and efficient operations of existing nuclear power plants and designing, licensing and deploying new reactor designs, however, will require focused R&D programs as well as the extensive use and leveraging of advanced modeling and simulation (M&S). M&S will play a key role in ensuring safe and efficient operations of existing and new nuclear reactors. The DOE-NE has been actively developing and promoting the use of advanced M&S in reactor design and analysis through its R&D programs, e.g., the Nuclear Energy Advanced Modeling and Simulation (NEAMS) and Consortium for Advanced Simulation of Light Water Reactors (CASL) programs. Also, nuclear reactor vendors are already using CFD and CSM, for design, analysis, and licensing. However, these M&S tools cannot be used with confidence for nuclear reactor applications unless accompanied and supported by verification and validation (V&V) and uncertainty quantification (UQ) processes and procedures which provide quantitative measures of uncertainty for specific applications. The Nuclear Energy Knowledge base for Advanced Modeling and Simulation (NE-KAMS) is being developed at the Idaho National Laboratory in conjunction with Bettis Laboratory, Sandia National Laboratories, Argonne National Laboratory, Utah State University and others with the objective of establishing a comprehensive and web-accessible knowledge base that will provide technical services and resources for V&V and UQ of M&S in nuclear energy sciences and engineering. The knowledge base will serve as an important resource for technical exchange and collaboration that will enable credible and reliable computational models and simulations for application to nuclear reactor design, analysis and licensing. NE-KAMS will serve as a valuable resource for the nuclear industry, academia, the national laboratories, the U.S. Nuclear Regulatory Commission (NRC) and the public and will help ensure the safe, economical and reliable operation of existing and future nuclear reactors. From its inception, NE-KAMS will directly support nuclear energy research, development and demonstration programs within the U.S. Department of Energy (DOE), including the CASL, NEAMS, Light Water Reactor Sustainability (LWRS), Small Modular Reactors (SMR), and Next Generation Nuclear Power Plant (NGNP) programs. These programs all involve M&S of nuclear reactor systems, components and processes, and it is envisioned that NE-KAMS will help to coordinate and facilitate collaboration and sharing of resources and expertise for V&V and UQ across these programs.

  5. Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act

    SciTech Connect (OSTI)

    NONE

    1988-01-01T23:59:59.000Z

    The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended by the Secretary of Energy and approved by the President for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared by the US Department of Energy (DOE) in accordance with the requirements of the Nulcear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site;to describe the conceptual designs for the repository and the waste package;and to present the plans for obtaining the geologic information necessary to demonstrate the suitability of the site for a repository, to design the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and developing a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next;it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing principles, and organization of the site characterization plan are oulined, and compliance with applicable regulations is discussed.

  6. Microsoft PowerPoint - NRC Nuclear Export Controls Implementing the NSG Trigger List_Gary Langlie [Compatibility Mode]

    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:AdministrationAnalysis and Feedback onWorkingIAEAMOXRole

  7. Appendix A U.S. Nuclear Waste Technical Review

    E-Print Network [OSTI]

    Appendices Appendices 37 #12;#12;Appendix A Appendix A U.S. Nuclear Waste Technical Review Board as chair, on the U.S. Nuclear Regulatory Commission's Advisory Commit tee on Nuclear Waste. His areas to the Nuclear Waste Technical Review Board on June 26, 2002, by President George W. Bush. Dr. Abkowitz

  8. Appendix A U.S. Nuclear Waste Technical Review

    E-Print Network [OSTI]

    Appendices Appendices 31 #12;#12;Appendix A Appendix A U.S. Nuclear Waste Technical Review Board.S. Nuclear Waste Technical Review Board as Chairman on September 10, 2004, by President George W. Bush. Dr­2004), 4 years as chair, on the U.S. Nuclear Regulatory Commission's Advisory Committee on Nuclear Waste

  9. RESEARCH PAPER Composition of the plant nuclear envelope: theme and

    E-Print Network [OSTI]

    Meier, Iris

    RESEARCH PAPER Composition of the plant nuclear envelope: theme and variations Iris Meier* Plant plants is only just beginning, fundamental differences from the animal nuclear envelope have already been to known plant regulatory pathways. Plant nuclear envelope composition The inner nuclear envelope A number

  10. Analysis of transcriptional regulatory circuitry

    E-Print Network [OSTI]

    Rinaldi, Nicola J., 1974-

    2004-01-01T23:59:59.000Z

    The research in this thesis has focused on the analysis of data from two types of microarray technologies with the goal of improving understanding of transcriptional regulatory circuitry in yeast. These microarray technologies, ...

  11. Proceedings of GLOBAL 2013: International Nuclear Fuel Cycle Conference - Nuclear Energy at a Crossroads

    SciTech Connect (OSTI)

    NONE

    2013-07-01T23:59:59.000Z

    The Global conference is a forum for the discussion of the scientific, technical, social and regulatory aspects of the nuclear fuel cycle. Relevant topics include global utilization of nuclear energy, current fuel cycle technologies, advanced reactors, advanced fuel cycles, nuclear nonproliferation and public acceptance.

  12. Regulatory Requirements for Cogeneration Projects

    E-Print Network [OSTI]

    Curry, K. A., Jr.

    1982-01-01T23:59:59.000Z

    for cogeneration, therefore, the discussion will be limited to those portions of each act that affect cogenerators. Since the original cogeneration legislation was passed in 1978 and implemented by the Federal Energy Regulatory Commission (FERC) in 1980... major pieces of legislation that impact cogeneration as well as an outline of the major provisions obtain ed in the Department of Energy Federal Energy Regulatory Commission final rule implementing Section 201 and Section 210 of PURPA. Public Uti...

  13. Regulatory facility guide for Ohio

    SciTech Connect (OSTI)

    Anderson, S.S.; Bock, R.E.; Francis, M.W.; Gove, R.M.; Johnson, P.E.; Kovac, F.M.; Mynatt, J.O. [Oak Ridge National Lab., TN (United States); Rymer, A.C. [Transportation Consulting Services, Knoxville, TN (United States)

    1994-02-28T23:59:59.000Z

    The Regulatory Facility Guide (RFG) has been developed for the DOE and contractor facilities located in the state of Ohio. It provides detailed compilations of international, federal, and state transportation-related regulations applicable to shipments originating at destined to Ohio facilities. This RFG was developed as an additional resource tool for use both by traffic managers who must ensure that transportation operations are in full compliance with all applicable regulatory requirements and by oversight personnel who must verify compliance activities.

  14. Regulatory Requirements | The Ames Laboratory

    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 PossibleRadiation Protection Technical s o Freiberg and SondershausenRegulatory DriversRegulatory

  15. Survey of thermal-hydraulic models of commercial nuclear power plants

    SciTech Connect (OSTI)

    Determan, J.C.; Hendrix, C.E.

    1992-12-01T23:59:59.000Z

    A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC's current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described.

  16. Survey of thermal-hydraulic models of commercial nuclear power plants

    SciTech Connect (OSTI)

    Determan, J.C.; Hendrix, C.E.

    1992-12-01T23:59:59.000Z

    A survey of the thermal-hydraulic models of nuclear power plants has been performed to identify the NRC`s current analytical capabilities for critical event response. The survey also supports ongoing research for accident management. The results of the survey are presented here. The PC database which records detailed data on each model is described.

  17. THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE

    E-Print Network [OSTI]

    Wang, J.S.Y.

    2010-01-01T23:59:59.000Z

    f o r 1 0 - y e a r - o l d PWR SF. I n i t i a l Heat G e nt nuclear fuel cycles for a PWR. Decay heat power for d i fKBS LWR MOX NRC NWTS ONWI OWI PWR RH-TRU WIPP b o i l i n g

  18. Threatened and endangered species evaluation for 75 licensed commercial nuclear power generating plants

    SciTech Connect (OSTI)

    Sackschewsky, M.R.

    1997-03-01T23:59:59.000Z

    The Endangered Species Act (ESA) of 1973, as amended, and related implementing regulations of the jurisdictional federal agencies, the U.S. Departments of Commerce and Interior, at 50 CFR Part 17. 1, et seq., require that federal agencies ensure that any action authorized, funded, or carried out under their jurisdiction is not likely to jeopardize the continued existence of any threatened or endangered species or result in the destruction or adverse modification of critical habitats for such species. The issuance and maintenance of a federal license, such as a construction permit or operating license issued by the U.S. Nuclear Regulatory Commission (NRC) for a commercial nuclear power generating facility is a federal action under the jurisdiction of a federal agency, and is therefore subject to the provisions of the ESA. The U.S. Department of the Interior (through the Fish and Wildlife Service), and the U.S. Department of Commerce, share responsibility for administration of the ESA. The National Marine Fisheries Service (NMFS) deals with species that inhabit marine environments and anadromous fish, while the U.S. Fish and Wildlife Service (USFWS) is responsible for terrestrial and freshwater species and migratory birds. A species (or other distinct taxonomic unit such as subspecies, variety, and for vertebrates, distinct population units) may be classified for protection as `endangered` when it is in danger of extinction within the foreseeable future throughout all or a significant portion of its range. A `threatened` classification is provided to those animals and plants likely to become endangered within the foreseeable future throughout all or a significant portion of their ranges. As of February 1997, there were about 1067 species listed under the ESA in the United States. Additionally there were approximately 125 species currently proposed for listing as threatened or endangered, and another 183 species considered to be candidates for formal listing proposals.

  19. An Empirical Study on Ultrasonic Testing in Lieu of Radiography for Nuclear Power Plants

    SciTech Connect (OSTI)

    Moran, Traci L.; Pardini, Allan F.; Ramuhalli, Pradeep; Prowant, Matthew S.; Mathews, Royce

    2012-09-01T23:59:59.000Z

    Research is being conducted for the U.S. Nuclear Regulatory Commission (NRC) at the Pacific Northwest National Laboratory (PNNL) to assess the capability, effectiveness, and reliability of ultrasonic testing (UT) as a replacement method for radiographic testing (RT) for inspecting nuclear power plant (NPP) components. A primary objective of this work is to evaluate UT techniques to assess their ability to detect, locate, size, and characterize fabrication flaws in typical NPP weldments. This particular study focused on the evaluation of four carbon steel pipe-to-pipe welds on specimens that ranged in thicknesses from 19.05 mm (0.75 in.) to 27.8 mm (1.094 in.) and were 355.6 mm (14.0 in.) or 406.4 mm (16.0 in.) in diameter. The pipe welds contained both implanted (intentional) fabrication flaws as well as bonus (unintentional) flaws throughout the entire thickness of the weld and the adjacent base material. The fabrication flaws were a combination of planar and volumetric flaw types, including incomplete fusion, incomplete penetration, cracks, porosity, and slag inclusions. The examinations were conducted using phased-array UT (PA UT) techniques applied primarily for detection and length sizing of the flaws. Radiographic examinations were also conducted on the specimens with RT detection and length sizing results being used to establish true state. This paper will discuss the comparison of UT and RT (true state) detection results conducted to date along with a discussion on the technical gaps that need to be addressed before these methods can be used interchangeably for repair and replacement activities for NPP components.

  20. Extra-regulatory impact tests and analyses of the structural evaluation test unit

    SciTech Connect (OSTI)

    Ludwigsen, J.S.; Ammerman, D.J.

    1995-05-01T23:59:59.000Z

    The structural evaluation test unit is roughly equivalent to a 1/3 scale model of a high level waste rail cask. The test unit was designed to just meet the requirements of NRC Regulatory Guide 7.6 when subjected to a 9 m (30 ft) free drop resulting in an impact velocity of 13.4 m/s (30 mph) onto an unyielding target in the end-on orientation. The test unit was then subjected to impacts with higher velocities to determine the amount of built-in conservatism in this design approach. Test impacts of 13.4, 20.1 and 26.8 m/s (30, 45, and 60 mph) were performed. This paper will describe the design, testing, and comparison of measured strains and deformations to the equivalent analytical predictions.

  1. Licensed fuel facility status report: Inventory difference data, July 1, 1994--June 30, 1995. Volume 15

    SciTech Connect (OSTI)

    Joy, D.R.

    1996-05-01T23:59:59.000Z

    The Nuclear Regulatory Commission (NRC) is committed to the periodic publication of licensed fuel facility inventory difference data, following agency review of the information and completion of any related NRC investigations. Information in this report includes inventory difference data for active fuel fabrication facilities possessing more than one effective kilogram of special nuclear material.

  2. Appendix F References | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    December. Washington, D.C. Available online at: http:www.nrc.govreading-rmdoc-collectionsnuregsstaffsr1811finalsr1811.html (accessed May 29, 2012) NRC 2006 Saricks,...

  3. Fresh nuclear fuel measurements at Ukrainian nuclear power plants

    SciTech Connect (OSTI)

    Kuzminski, Jozef [Los Alamos National Laboratory; Ewing, Tom [ANL; Dickman, Debbie [PNNL; Gavrilyuk, Victor [UKRAINE; Drapey, Sergey [UKRAINE; Kirischuk, Vladimir [UKRAINE; Strilchuk, Nikolay [UKRAINE

    2009-01-01T23:59:59.000Z

    In 2005, the Provisions on Nuclear Material Measurement System was enacted in Ukraine as an important regulatory driver to support international obligations in nuclear safeguards and nonproliferation. It defines key provisions and requirements for material measurement and measurement control programs to ensure the quality and reliability of measurement data within the framework of the State MC&A System. Implementing the Provisions requires establishing a number of measurement techniques for both fresh and spent nuclear fuel for various types of Ukrainian reactors. Our first efforts focused on measurements of fresh nuclear fuel from a WWR-1000 power reactor.

  4. Analysis of a CRDM Nozzle Break LOCA Without Scram Using the U.S. NRC Coupled Code TRAC-M/PARCS

    SciTech Connect (OSTI)

    Cuadra, A.; Ragusa, J.; Downar, T. [Purdue University, West Lafayette, IN 47907 (United States); Ivanov, K. [The Pennsylvania State University, University Park, PA 16802 (United States)

    2002-07-01T23:59:59.000Z

    Recently, through-wall circumferential cracks in several control rod drive mechanism (CRDM) nozzle penetrations were detected at the Oconee-3 nuclear power plant. The presence of these cracks was seen as a potential precursor to a small break loss of coolant accident. In order to assess the impact of a postulated failure of a CRDM housing, analyses were performed using the U.S. NRC coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. Although deemed highly unlikely, it was assumed that no control rods inserted in order to bound any possible reactivity transient associated with the break. The thermal-hydraulic model used to perform the study is based on an existing model of Oconee, built for PTS analysis, which models the whole plant and some of its control systems. A refined vessel model based on a TMI model was used to increase the resolution of the results and facilitate coupling to PARCS. All relevant ECCS systems were modeled and the control system allowed for, in addition to automatic actions, some assumed operator intervention. In particular, HPI throttling was modeled to maintain the hot leg subcooled at 42 +/- 7 K, and prevent an excessive amount of cold water from being injected into the system. A spatial kinetics analysis of this event was necessary because of the wide range of core conditions which occurred during the transient, from hot full power operation conditions to the cold zero power shutdown state. Analysis of the event with point kinetics and 'best estimate' reactivity coefficients resulted in significant miss-prediction of the core power response. Conversely, the three-dimensional kinetics solution with cross section data generated over the entire range of the event led to a more accurate calculation of the power response and the overall analysis of the system transient response. This paper will describe the analysis of the control rod drive nozzle break event without scram using TRAC-M/PARCS. (authors)

  5. Overview of ORNL/NRC programs addressing durability of concrete structures

    SciTech Connect (OSTI)

    Naus, D.J.; Oland, C.B.

    1994-06-01T23:59:59.000Z

    The role of reinforced concrete relative to its applications as either safety-related structures in nuclear power or engineered barriers of low-level radioactive waste disposal facilities is described. Factors that can affect the long-term durability of reinforced concrete are identified. Overviews are presented of the Structural Aging Program, which is addressing the aging management of safety-related concrete structures in nuclear power plants, and the Permeability Test Methods and Data Program, which is identifying pertinent data and information for use in performance assessments of engineered barriers for low-level radioactive waste disposal.

  6. THE NRC HAS ISSUED Ati ACTION PLAN FOR CLEANUP OF 46 SfTES that;;ece

    Office of Legacy Management (LM)

    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 group currentBradleyTableSelling7 AugustAFRICAN3uj: ;;I : T' j-jE: iLie'Al ...NRC

  7. Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS) Code Verification and Validation Data Standards and Requirements: Fluid Dynamics Version 1.0

    SciTech Connect (OSTI)

    Greg Weirs; Hyung Lee

    2011-09-01T23:59:59.000Z

    V&V and UQ are the primary means to assess the accuracy and reliability of M&S and, hence, to establish confidence in M&S. Though other industries are establishing standards and requirements for the performance of V&V and UQ, at present, the nuclear industry has not established such standards or requirements. However, the nuclear industry is beginning to recognize that such standards are needed and that the resources needed to support V&V and UQ will be very significant. In fact, no single organization has sufficient resources or expertise required to organize, conduct and maintain a comprehensive V&V and UQ program. What is needed is a systematic and standardized approach to establish and provide V&V and UQ resources at a national or even international level, with a consortium of partners from government, academia and industry. Specifically, what is needed is a structured and cost-effective knowledge base that collects, evaluates and stores verification and validation data, and shows how it can be used to perform V&V and UQ, leveraging collaboration and sharing of resources to support existing engineering and licensing procedures as well as science-based V&V and UQ processes. The Nuclear Energy Knowledge base for Advanced Modeling and Simulation (NE-KAMS) is being developed at the Idaho National Laboratory in conjunction with Bettis Laboratory, Sandia National Laboratories, Argonne National Laboratory, Utah State University and others with the objective of establishing a comprehensive and web-accessible knowledge base to provide V&V and UQ resources for M&S for nuclear reactor design, analysis and licensing. The knowledge base will serve as an important resource for technical exchange and collaboration that will enable credible and reliable computational models and simulations for application to nuclear power. NE-KAMS will serve as a valuable resource for the nuclear industry, academia, the national laboratories, the U.S. Nuclear Regulatory Commission (NRC) and the public and will help ensure the safe, economical and reliable operation of existing and future nuclear reactors.

  8. Nuclear Material Control and Accountability

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

    2011-06-27T23:59:59.000Z

    This Order establishes performance objectives, metrics, and requirements for developing, implementing, and maintaining a nuclear material control and accountability program within DOE/NNSA and for DOE-owned materials at other facilities that are exempt from licensing by the Nuclear Regulatory Commission. Cancels DOE M 470.4-6, Admin Chg 1, 8-26-05. Admin Chg 2, dated 11-19-12, cancels DOE M 474.2 Admin Chg 1.

  9. Questions and answers based on revised 10 CFR Part 20

    SciTech Connect (OSTI)

    Borges, T.; Stafford, R.S.; Lu, P.Y. [Oak Ridge National Lab., TN (United States); Carter, D. [Nuclear Regulatory Commission, Washington, DC (United States)

    1994-05-01T23:59:59.000Z

    NUREG/CR-6204 is a collection of questions and answers that were originally issued in seven sets and which pertain to revised 10 CFR Part 20. The questions came from both outside and within the NRC. The answers were compiled and provided by NRC staff within the offices of Nuclear Reactor Regulation, Nuclear Material Safety and Safeguards, Nuclear Regulatory Research, the Office of State Programs, and the five regional offices. Although all of the questions and answers have been reviewed by attorneys in the NRC Office of the General Counsel, they do not constitute official legal interpretations relevant to revised 10 CFR Part 20. The questions and answers do, however, reflect NRC staff decisions and technical options on aspects of the revised 10 CFR Part 20 regulatory requirements. This NUREG is being made available to encourage communication among the public, industry, and NRC staff concerning the major revisions of the NRC`s standards for protection against radiation.

  10. Threat Insight Quarterly Regulatory Compliance

    E-Print Network [OSTI]

    X-Force ® Threat Insight Quarterly Regulatory Compliance www.iss.netwww.iss.net October 2006 #12 Risk Index ..............................................11 Future X-Force Threat Insight Quarterly Internet Security Systems X-Force® Threat Insight Quarterly > October 2006 ©2006 Internet Security Systems

  11. State Regulatory Oversight of Geothermal

    E-Print Network [OSTI]

    State Regulatory Oversight of Geothermal Heat Pump Installations: 2012 Kevin McCray Executive of this project was to update previous research accomplished by the Geothermal Heat Pump Consortium (GHPC of ground-source heat pump (GSHP) systems. The work was to provide insight into existing and anticipated

  12. Entry/exit control at fuel fabrication facilities using or possessing formula quantities of strategic special nuclear material

    SciTech Connect (OSTI)

    Dwyer, P.A.

    1988-12-01T23:59:59.000Z

    This document presents information on entry/exit control at fuel fabrication facilities using or possessing formula quantities of strategic special nuclear material. It describes NRC requirements and methods for conducting personnel, package, and vehicle searches at these facilities. Testing methods for determining the detection capability of firearms, explosives, and metal detectors are provided.

  13. State Regulatory Oversight of Geothermal Heat Pump

    E-Print Network [OSTI]

    State Regulatory Oversight of Geothermal Heat Pump Installa:ons: 2012 & 2009 Kevin McCray, Execu:ve Director #12;2009 #12;Sponsors ·The Geothermal Hea requested geothermal hea:ng and cooling regulatory data. · An email containing

  14. Title list of documents made publicly available: September 1--30, 1996. Volume 18, Number 9

    SciTech Connect (OSTI)

    NONE

    1996-11-01T23:59:59.000Z

    The report describes the information received and published by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) non-docketed material received and published by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  15. Title list of documents made publicly available, May 1-31, 1996

    SciTech Connect (OSTI)

    Morris, E.B.

    1996-07-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It describes the information received and published by the U.S. Nuclear Regulatory Commission (NRC). This information includes (1) docketed, material associated with civilian nuclear power plants and other uses of radioactive materials and (2) non-docketed material received and published by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index and a Report Number Index.

  16. Title list of documents made publicly available: October 1--31, 1994. Volume 16, Number 10

    SciTech Connect (OSTI)

    Not Available

    1994-12-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It contains descriptions of the information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  17. Title list of documents made publicly available, January 1, 1997--January 31, 1997

    SciTech Connect (OSTI)

    NONE

    1997-04-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It describes the information received and published by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) non-docketed material received and published by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  18. Title list of documents made publicly available. Volume 16, Number 5

    SciTech Connect (OSTI)

    Not Available

    1994-07-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available contains descriptions of the information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  19. Title list of documents made publicly available: June 1--30, 1995. Volume 17, Number 6

    SciTech Connect (OSTI)

    NONE

    1995-08-01T23:59:59.000Z

    This monthly publication contains descriptions of the information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  20. Title list of documents made publicly available: April 1--30, 1996. Volume 18, Number 4

    SciTech Connect (OSTI)

    NONE

    1996-06-01T23:59:59.000Z

    This publication describes the information received and published by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) non-docketed material received and published by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  1. Title list of documents made publicly available: February 1--29, 1996. Volume 18, Number 2

    SciTech Connect (OSTI)

    NONE

    1996-04-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It contains descriptions of the information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  2. Title list of documents made publicly available, March 1--31, 1998. Volume 20, Number 3

    SciTech Connect (OSTI)

    NONE

    1998-05-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It describes the information received and published by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and published by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a personal author index, a corporate source index, and a report number index.

  3. Title list of documents made publicly available: May 1--31, 1997. Volume 19, Number 5

    SciTech Connect (OSTI)

    NONE

    1997-07-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It describes the information received and published by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) non-docketed material received and published by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  4. Title list of documents made publicly available, September 1-30, 1997

    SciTech Connect (OSTI)

    NONE

    1997-11-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It describes the information received and published by the U.S. Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and published by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  5. Title list of documents made publicly available, December 1--31, 1993, Volume 15, No. 12

    SciTech Connect (OSTI)

    NONE

    1994-02-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It contains descriptions of the information received and generated by the U.S. Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  6. Title list of documents made publicly available. Volume 17, No. 5

    SciTech Connect (OSTI)

    NONE

    1995-07-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It contains descriptions of the information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (3) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  7. Title list of documents made publicly available. Volume 17, No. 7

    SciTech Connect (OSTI)

    NONE

    1995-09-01T23:59:59.000Z

    The Title List of Documents Made Publicly Available is a monthly publication. It contains descriptions of the information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. As used here, docketed does not refer to Court dockets; it refers to the system by which NRC maintains its regulatory records. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  8. Confirmatory Survey Results for the Emergency Operations Facility (EOF) at the Connecticut Yankee Haddam Neck Plant, Haddam, Connecticut

    SciTech Connect (OSTI)

    W. C. Adams

    2007-07-03T23:59:59.000Z

    The U.S. Nuclear Regulatory Commission (NRC) requested that the Oak Ridge Institute for Science and Education (ORISE) perform a confirmatory survey on the Emergency Operations Facility (EOF) at the Connecticut Yankee Haddam Neck Plant (HNP) in Haddam, Connecticut

  9. Managing Tipping Point Dynamics in Complex Construction Projects

    E-Print Network [OSTI]

    Ford, David N.

    -9364 2008 134:6 421 CE Database subject headings: Project management; Dynamic models; Simulation models from the Nuclear Regulatory Commission NRC 1982 and the Energy In- formation Administration EIA 1988

  10. Appendix V. List of Acronyms and Abbreviations, and Glossary of Terms List of Acronyms and Abbreviations

    E-Print Network [OSTI]

    of Radiation Control Program Directors Cu copper CWA Clean Water Act D&D decontamination and decommissioning Service NRC Nuclear Regulatory Commission NSPS New Source Performance Standards O2 oxygen ORIA Office

  11. Appendix I. List of Acronyms and Abbreviations, and Glossary of Terms

    E-Print Network [OSTI]

    Control Program Directors Cu copper CWA Clean Water Act D&D decontamination and decommissioning DOE Priorities List NPS National Park Service NRC Nuclear Regulatory Commission NSPS New Source Performance

  12. NUREG-1575, Supp. 1 EPA 402-R-09-001

    E-Print Network [OSTI]

    .................................................................................... xxvii Conversion Factors of Materials and Equipment Manual (MARSAME) Department of Defense Department of Energy Environmental Protection of Energy (DOE), Environmental Protection Agency (EPA), and Nuclear Regulatory Commission (NRC

  13. Uncertainty and sensitivity analysis of a fire-induced accident scenario involving binary variables and mechanistic codes

    E-Print Network [OSTI]

    Minton, Mark A. (Mark Aaron)

    2010-01-01T23:59:59.000Z

    In response to the transition by the United States Nuclear Regulatory Commission (NRC) to a risk-informed, performance-based fire protection rulemaking standard, Fire Probabilistic Risk Assessment (PRA) methods have been ...

  14. Simulation of condensation systems in the presence of noncondensable gases

    E-Print Network [OSTI]

    Raja, Laxminarayan Lakshmana

    1992-01-01T23:59:59.000Z

    in a, pressurized water reactor. I. 4 RELAP5 Computer Code The RELAP5 computer code is a, light water transient analysis code developed for the U. S. Nuclear Regulatory Commission (NRC) for use in rule making, licensing audit calculations...

  15. Comment and response document for the final remedial action plan site design for stabilization of the Inactive Uranium Mill Tailings Sites at Slick Rock, Colorado

    SciTech Connect (OSTI)

    NONE

    1995-09-01T23:59:59.000Z

    This document consists of comments and responses; the reviewers are the U.S. Nuclear Regulatory Commission (NRC), Colorado Dept. of Public Health and Environment, and the remedial action contractor (RAC).

  16. Comparison of Average Transport and Dispersion Among a Gaussian, a Two-Dimensional, and a Three-Dimensional Model

    Broader source: Energy.gov [DOE]

    The Nuclear Regulatory Commission’s (NRC’s) code for predicting off-siteconsequences, MACCS2 (Chanin, et al. 1998) (MELCOR Accident Consequence CodeSystem, Version 2), uses a simplified model for...

  17. Moving Toward Product Line Engineering in a Nuclear Industry Consortium

    E-Print Network [OSTI]

    Boyer, Edmond

    power have special institutions overseeing and regulating nuclear safety. Nuclear industry projects must conform to na- tional safety institutions and international regulations. In many cases, regulatory sophisticated and complex energy systems ever designed. Nuclear safety Permission to make digital or hard copies

  18. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    SciTech Connect (OSTI)

    Lee, Deokjung [Purdue University (United States); Downar, Thomas J. [Purdue University (United States); Ulses, Anthony [U.S. Nuclear Regulatory Commission (United States); Akdeniz, Bedirhan [Pennsylvania State University (United States); Ivanov, Kostadin N. [Pennsylvania State University (United States)

    2004-10-15T23:59:59.000Z

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect.

  19. Identification and Characterization of Prokaryotic Regulatory Networks: Final Report

    SciTech Connect (OSTI)

    Gary D Stormo

    2012-10-05T23:59:59.000Z

    We have completed our characterization of both the transcriptional regulatory network and post-transcriptional regulatory motifs in Shewanella.

  20. Fact Sheet: DOE/National Association of Regulatory Utility Commissione...

    Office of Environmental Management (EM)

    DOENational Association of Regulatory Utility Commissioners Natural Gas Infrastructure Modernization Partnership Fact Sheet: DOENational Association of Regulatory Utility...

  1. Control and Accountability of Nuclear Materials

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

    1994-09-07T23:59:59.000Z

    To prescribe the Department of Energy (DOE) minimum requirements and procedures for control and accountability of nuclear materials at DOE-owned and -leased facilities and DOE-owned nuclear materials at other facilities which are exempt from licensing by the Nuclear Regulatory Commission. Cancels DOE O 5633.2A and DOE O 5633.3A. Canceled by DOE O 474.1

  2. The U.S. regulatory framework for long-term management of uranium mill tailings

    SciTech Connect (OSTI)

    Smythe, C. [Dept. of Energy, Albuquerque, NM (United States); Bierley, D.; Bradshaw, M. [Roy F. Weston, Inc., Albuquerque, NM (United States)

    1995-03-01T23:59:59.000Z

    The US established the regulatory structure for the management, disposal, and long-term care of uranium mill tailings in 1978 with the passage of the Uranium Mill Tailings Radiation Control Act (UMTRCA) (Pub. L. 95-604). This legislation has governed the cleanup and disposal of uranium tailings at both inactive and active sites. The passage of the UMTRCA established a federal regulatory program for the cleanup and disposal of uranium mill tailings in the US. This program involves the DOE, the NRC, the EPA, various states and tribal governments, private licensees, and the general public. The DOE has completed surface remediation at 14 sites, with the remaining sites either under construction or in planning. The DOE`s UMTRA Project has been very successful in dealing with public and agency demands, particularly regarding disposal site selection and transportation issues. The active sites are also being cleaned up, but at a slower pace than the inactive sites, with the first site tentatively scheduled for completion in 1996.

  3. Title list of documents made publicly available. Volume 17, No. 10

    SciTech Connect (OSTI)

    NONE

    1995-12-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  4. Title list of documents made publicly available: August 1--31, 1997. Volume 19, Number 8

    SciTech Connect (OSTI)

    NONE

    1997-10-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  5. Title list of documents made publicly available, September 1--30, 1994. Volume 16, No. 9

    SciTech Connect (OSTI)

    Not Available

    1994-11-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  6. Title List of documents made publicly available, September 1--30, 1993. Volume 15, No. 9

    SciTech Connect (OSTI)

    Not Available

    1993-11-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principals Documents.

  7. Title list of documents made publicly available. Volume 17, No. 1

    SciTech Connect (OSTI)

    NONE

    1995-03-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal author, corporate source, report number, and cross reference of enclosures to principal documents.

  8. Title list of documents made publicly available, December 1--31, 1991. Volume 13, No. 12

    SciTech Connect (OSTI)

    None

    1992-02-01T23:59:59.000Z

    This document is a monthly containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  9. Title list of documents made publicly available, November 1--30, 1994. Volume 16, No. 11

    SciTech Connect (OSTI)

    Not Available

    1995-01-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  10. Title list of documents made publicly available: December 1--31, 1996. Volume 18, Number 12

    SciTech Connect (OSTI)

    NONE

    1997-03-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  11. Title list of documents made publicly available: November 1--30, 1997. Volume 19, Number 11

    SciTech Connect (OSTI)

    NONE

    1998-01-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  12. Title list of documents made publicly available, August 1--31, 1993. Volume 15, No. 8

    SciTech Connect (OSTI)

    Not Available

    1993-10-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory commission (NRC). This information includes docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  13. Title list of documents made publicly available, December 1--31, 1995. Volume 17, No. 12

    SciTech Connect (OSTI)

    NONE

    1996-02-01T23:59:59.000Z

    This document is a monthly publication containing descriptions of information received and generated by the US Nuclear Regulatory Commission (NRC). This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Personal Author, Corporate Source, Report Number, and Cross Reference of Enclosures to Principal Documents.

  14. Title of documents made publicly available, August 1--31, 1996

    SciTech Connect (OSTI)

    Yanez, V.E.

    1996-10-01T23:59:59.000Z

    This publication describes the information received and published by the US Nuclear Regulatory Commission (NRC). This information includes: (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) non-docketed material received and published by NRC pertinent to its role as a regulatory agency. This series of documents is indexed by a Personal Author Index, a Corporate Source Index, and a Report Number Index.

  15. Title list of documents made publicly available, December 1-31, 1997

    SciTech Connect (OSTI)

    NONE

    1998-02-01T23:59:59.000Z

    This report includes the information received and published by the U.S. Nuclear Regulatory Commission (NRC) in December 1997. It includes: (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) non-docketed material received and published by NRC pertinent to its role as a regulatory agency. This series of documents is indexed by author, corporate source, and report number. This report includes six docketed items and 24 non-docketed items.

  16. Microsoft Word - SmartGrid - NRC Input to DOE Requestrvjcomments.docx

    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 Questions forCheneyNovemberi CONTENTS ExecutiveHouse Committee on345ThisNuclear

  17. 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

  18. Title list of documents made publicly available, March 1-31, 1996

    SciTech Connect (OSTI)

    NONE

    1996-05-01T23:59:59.000Z

    This document is the March 1996 listing of publicly available documents published by the NRC. The documents include: (1) docketed material associated with civilian nuclear power plants and other uses of radioactive material, and (2) nondocketed material received and generated by the NRC in its role as a regulatory agency.

  19. Title list of documents made publicly available, July 1--31, 1996: Volume 18, No. 7

    SciTech Connect (OSTI)

    NONE

    1996-09-01T23:59:59.000Z

    This monthly publication describes information received and published by US NRC. This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials and (2) nondocketed material received and published by NRC pertinent to its role as a regulatory agency. Following indexes are included: personal author, corporate source, report number, and cross reference of enclosures to principal documents.

  20. Designation Order No. 00-09.00 to the Under Secretary

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

    2006-05-12T23:59:59.000Z

    Designated as the authorized representative to perform on the Secretary of Energy's behalf as the "authorized representative", as that term is used in the Nuclear Regulatory Commission (NRC) regulations 10 CFR 72.16(b), and as named holder of the NRC license under 10 CFR Part 72 for Idaho Spent Fuel Facility Independent Spent Fuel Storage Installation.

  1. Redesignation Order No. 00-09.01 to the Assistant Secretary for Environmental Management

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

    2006-05-12T23:59:59.000Z

    Redesignation of the Assistant Secretary for Environmental Management of DOE as the authorized representative to perform as the "Authorized Representative," as that term is used in the Nuclear Regulatory Commission (NRC) regulations at 10 CFR 72.16(b) and as name holder of the NRC license under 10 CFR Part 72 of for the Idaho Spent Fuel Facility Independent Spent Fuel Storage Installation.

  2. Budget estimates fiscal year 1995: Volume 10

    SciTech Connect (OSTI)

    Not Available

    1994-02-01T23:59:59.000Z

    This report contains the Nuclear Regulatory Commission (NRC) fiscal year budget justification to Congress. The budget provides estimates for salaries and expenses and for the Office of the Inspector General for fiscal year 1995. The NRC 1995 budget request is $546,497,000. This is an increase of $11,497,000 above the proposed level for FY 1994. The NRC FY 1995 budget request is 3,218 FTEs. This is a decrease of 75 FTEs below the 1994 proposed level.

  3. Regulatory Processes | 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 Strategic2 OPAM615_CostNSAR -Department of Energyas ofCheniereDepartment ofRegulatory

  4. Nuclear Regulatory Commission issuances. Volume 38, No. 2

    SciTech Connect (OSTI)

    Not Available

    1993-08-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors; Decisions (DD), and the Denials of Petitions for Rulemaking (DPRM).

  5. Nuclear Regulatory Commission issuances. Volume 47, Number 3

    SciTech Connect (OSTI)

    NONE

    1998-03-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM). The two issuances included here are: (1) the Commission issuance to the US Enrichment Corporation and (2) the Director`s Decision to the North Atlantic Energy Service Corporation.

  6. Nuclear Regulatory Commission Issuances. Volume 45, No. 3

    SciTech Connect (OSTI)

    NONE

    1997-03-01T23:59:59.000Z

    This report includes the issuances received in March 1997. Issuances are from the Commission, the Atomic Safety and Licensing Boards, and the Directors` Decisions. 10 issuances were received: Louisiana Energy Services (2 issuances); Illinois Power Company and Soyland Power Cooperative; Ralph. L. Tetrick; University of Cincinnati; Consumers Power Company; Entergy Operations, Inc.; Georgia Power Company; Westinghouse Electric Corporation; and Wisconsin Electric Power Company. No issuances were received from the the Administrative Law Judges or the Decisions on Petitions for Rulemaking.

  7. Nuclear Regulatory Commission issuances, April 1994. Volume 39, No. 4

    SciTech Connect (OSTI)

    Not Available

    1994-04-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Denials of Petitions for Rulemaking (DPRM).

  8. Nuclear Regulatory Commission Issuances, Volume 44, No. 5

    SciTech Connect (OSTI)

    NONE

    1996-11-01T23:59:59.000Z

    This report includes the issuances received in November 1996. Issuances are from the Commission, the Atomic Safety and Licensing Boards, and the Directors` Decisions. Seven issuances were received and are abstracted individually in the database: Emerick S. McDaniel, U.S. Enrichment Corporation, Sequoyah Fuels Corporation and General Atomics, all power reactor licensees, Florida Power and Light Company, Maine Yankee Atomic Power Company, and Northern States Power Company. No issuances were received from the the Administrative Law Judges or the Decisions on Petitions for Rulemaking.

  9. Nuclear Regulatory Commission issuances: Volume 46, Number 6

    SciTech Connect (OSTI)

    NONE

    1997-12-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decision (DD), and the Decision on Petitions for Rulemaking (DPRM). The summaries and headnotes preceding the opinions reported herein are not to be deemed a part of those opinions or have any independent legal significance.

  10. UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C...

    Office of Scientific and Technical Information (OSTI)

    . These releases, t h e c0ntainmen.t f a i l u r e modes, and t h e t i m i n g t o a computer code c a l l e d CORRAL which c a l c u l a t e s t h e amounts o f t h e v a r i o u...

  11. Nuclear Regulatory Commission issuances. Volume 47, Number 2

    SciTech Connect (OSTI)

    NONE

    1998-02-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM). They involve: 21st Century Technologies, Inc.; Hydro Resources, Inc.; Northeast Utilities; and Northern States Power Company.

  12. Nuclear Regulatory Commission issuances, February 1996: Volume 43, No. 2

    SciTech Connect (OSTI)

    NONE

    1996-02-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM).

  13. Nuclear Regulatory Commission issuances. Volume 46, Number 1

    SciTech Connect (OSTI)

    NONE

    1997-07-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM). The summaries and headnotes preceding the opinions reported herein are not to be deemed a part of those opinions or have any independent legal significance.

  14. Nuclear Regulatory Commission issuances: June 1995. Volume 41, Number 6

    SciTech Connect (OSTI)

    NONE

    1995-06-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Decision on Petitions for Rulemaking (DPRM). This report contains four issuances by the CLI, 5 issuances by the LBP, and 1 issuance by the DD.

  15. Indexes to Nuclear Regulatory Commission Issuances, January--June 1997

    SciTech Connect (OSTI)

    NONE

    1997-10-01T23:59:59.000Z

    This report indexes the issuances heard and ruled upon from January through June 1997. Issuances are from the Commission, the Atomic Safety and Licensing Boards, the Administrative Law Judges, the Directors` Decisions, and the Decisions on Petitions for Rulemaking. Information provided for each case includes the case name, full test reference, issuance number, issues raised by appellants, legal citations, facility name and docket number, subject matter of issues and/or rulings, type of hearing, and type of issuance. Each issuance is displayed in one or more of five separate formats: (1) case name index, (2) headers and digests, (3) legal citations index, (4) subject index, and (5) facility index.

  16. Energy Praises the Nuclear Regulatory Commission Approval of the First

    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 YouTube|6721 FederalTexas Energy Incentive Programs,EnergyAugustPublic Law of theUnited

  17. Energy Department Issues Draft Request for Proposal for Nuclear Regulatory

    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 742EnergyOnItem NotEnergy,ARMFormsGasRelease Date:researchEmerging ThreatsEmploymentto

  18. Nuclear Regulatory Commission issuances. Volume 42, No. 4

    SciTech Connect (OSTI)

    NONE

    1995-10-01T23:59:59.000Z

    This report includes the issuances received during the specified period from the Commission (CLI), the Atomic Safety and Licensing Boards (LBP), the Administrative Law Judges (ALJ), the Directors` Decisions (DD), and the Decisions on Petitions for Rulemaking (DPRM).

  19. UNITED STATES NUCLEAR REGULATORY COMMISSION WAWINGTON. D. C. 20555

    Office of Legacy Management (LM)

    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 group currentBradleyTableSelling7 AugustAFRICAN3uj: ;;IDEC. i' ,' ,'WAWINGTON. D.

  20. UNITED STATES NUCLEAR REGULATORY COMMISSION WAWINQTON, 0. C. ZOSSS

    Office of Legacy Management (LM)

    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 group currentBradleyTableSelling7 AugustAFRICAN3uj: ;;IDEC. i' ,' ,'WAWINGTON.