Bøger af U S Nuclear Regulatory Commission

The Standard Review Plan (SRP) for dry storage systems (DSS) provides guidance to the U.S. Nuclear Regulatory Commission (NRC) staff in the Division of Spent Fuel Storage and Transportation (SFST) for reviewing applications for a Certificate of Compliance (CoC) of a dry storage system (DSS) for use at a general license facility.

The NRC is establishing a risk-informed revision of the design-basis pipe break size requirements in 10 CFR 50.46, Appendix K to Part 50, and GDC 35 which requires estimates of LOCA frequencies as a function of break size. Separate BWR and PWR piping and non-piping passive system LOCA frequency estimates were developed as a function of effective break size and operating time through the end of the plant license-renewal period. The estimates were based on an expert elicitation process which consolidated operating experience and insights from probabilistic fracture mechanics studies with knowledge of plant design, operation, and material performance. The elicitation required each member of an expert panel to qualitatively and quantitatively assess important LOCA contributing factors and quantify their uncertainty. The quantitative responses were combined to develop BWR and PWR total LOCA frequency estimates for each contributing panelist. The distributions for the six LOCA size categories and three time periods evaluated are represented by four parameters (mean, median, 5th and 95th percentiles). Finally, the individual estimates were aggregated to obtain group estimates, along with measures of panel diversity.

This safety evaluation report, documents the technical review of the Palisades Nuclear Plant license renewal application by the U.S. Nuclear Regulatory Commission Staff.

In September 2003, U.S. Nuclear Regulatory Commission (NRC) staff in the Office of Nuclear Material Safety and Safeguards (NMSS)1 consolidated and updated the policies and guidance of its decommissioning program in a three-volume NUREG series, NUREG-1757, "Consolidated NMSS Decommissioning Guidance." This NUREG series provides guidance on: planning and implementing license termination under the NRC's License Termination Rule (LTR), in the Code of Federal Regulations (CFR), Title 10, Part 20, Subpart E; complying with the radiological criteria for license termination; and complying with the requirements for financial assurance and recordkeeping for decommissioning and timeliness in decommissioning of materials facilities. The staff periodically updates NUREG-1757, so that it reflects current NRC decommissioning policy.

In response to the 1975 Browns Ferry Nuclear Power Plant fire, the U.S. Nuclear Regulatory Commission (NRC) issued Appendix R to Title 10 of the Code of Federal Regulations Part 50. To support fire protection defense-in-depth 1- or 3-hour electric raceway fire barrier systems (ERFBS) were permitted for use as an acceptable method to protect electrical cables essential to fire protection safe shutdown capability. However, ERFBS were a new approach to fire barrier applications and as the initial installation of the ERFBS began, there was uncertainty regarding the ERFBS performance and definitive test standards for ERFBS qualification. Following review and research efforts, NRC resolved many concerns with ERFBS, including the fire resistance, ampacity derating, and seismic position retention. This report documents the history of various ERFBS and how U.S. commercial nuclear power plants use ERFBS for compliance. This report also documents the current state of the use of ERFBS and evaluates the effectiveness of these barriers in achieving adequate protection for nuclear power plants.

The U.S. Nuclear Regulatory Commission (NRC) has been considering revision of the regulatory requirements for the emergency core cooling system (ECCS), as set forth in Title 10, Section 50.46, of the Code of Federal Regulations (10 CFR 50.46); Appendix K to 10 CFR Part 50; and General Design Criterion (GDC) 35. In particular, those requirements state that the ECCS shall be sized to provide adequate makeup water to compensate for a break of the largest diameter pipe in the primary system [i.e., the so-called "double-ended guillotine break" (DEGB)]. Consequently, in order to risk-inform the break size, the staff of the NRC proposed the concept of transition break size (TBS). In addition, the NRC developed pipe break frequencies as a function of break size using an expert elicitation process for degradation-related pipe breaks in reactorcoolantsystemsoftypicalboiling-andpressurized-waterreactors. Thatelicitationfocusedon determining event frequencies that initiate by unisolable primary side failures that can be exacerbated by material degradation with age under normal operating conditions. The purpose of this study was to assess potential seismic effects on the postulated TBS, and to provide information to facilitate review and comment regarding the proposed risk-informed revision of ECCS requirements in 10 CFR 50.46. Thus, this report evaluates the seismic effects, using different approaches to evaluate flawed and unflawed piping, and indirect failures of other components and component supports that could lead to piping failure.

This standard review plan (SRP) provides guidance to the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation and Office of Nuclear Material Safety and Safeguards, on how to evaluate requests to release part of a power reactor facility or site for unrestricted use prior to NRC approval of the facility's license termination plan (LTP). This type of release is referred as a "partial site release" (PSR). The associated regulatory requirements are contained in Title 10, Section 50.83, "Release of Part of a Power Reactor Facility or Site for Unrestricted Use," of the Code of Federal Regulations (10 CFR 50.83). This SRP presents evaluation and acceptance criteria for the review of PSR applications submitted by licensees. In so doing, this SRP identifies technical areas for review, information to be submitted by licensees, regulatory requirements and guidance, and evaluation criteria for use in assessing compliance with the regulations. Although the main focus of this guidance is directed to the NRC staff's review process, this SRP presents information that licensees and applicants may find useful. Another important purpose of the SRP is to make information available so that the public has a better understanding of the staff's review process and criteria for rejecting or accepting such requests.

This report provides staff and stakeholders with an understanding of Enterprise Content Management (ECM), describes the U.S. Nuclear Regulatory Commission's (NRC's) planned ECM Program, and discusses how this ECM Program will provide critical support to the NRC's mission and streamline the work of staff and stakeholders within the NRC's regulatory environment. ECM is the ability of an organization to capture, manage, store, preserve, and deliver needed information so that it is available to the right people at the right time.

This volume is one of five volumes shown in the table below that comprise the Safety Evaluation Report (SER). Each volume is to be published separately as it is completed; however, the volume number may not be published in sequence (e.g., Volume 3 is anticipated to be published before Volume 2). The SER volume number and section number of chapters within a volume, are based on the Yucca Mountain Review Plan (YMRP)1 that the U.S. Nuclear Regulatory Commission (NRC) staff used to guide its review of the license application. Use of SER section numbers that correspond to the YMRP section numbers facilitated the NRC staff's writing of the SER, allowing the interested reader to easily find the applicable review methods and acceptance criteria within the YMRP. The following table provides the topics and SER sections for each volume. The table will help the reader locate SER section cross-references in each volume.

One of the great challenges of strategies for deliver- ing climate information is matching user expecta- tions with the capacity to deliver that information. When work began on the First National Assessment in the late 1990s, a partnership of federal agencies conducted regional workshops across the country to engage scientists and stakeholders in identifying the types of climate information needed for making decisions and the best processes and practices for delivering that information.

This supplemental environmental impact statement (SEIS) documents the U.S. Nuclear Regulatory Commission (NRC) staff's analysis and conclusions regarding the environmental impacts of constructing and operating two new nuclear units (Units 3 and 4) at the Vogtle Electric Generating Plant (VEGP) site near Waynesboro, Georgia, and the mitigation measures available for reducing or avoiding adverse environmental impacts.

Supplement 3, "Guidance for Protective Action Strategies," to NUREG-0654/FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," updates the previous version of Supplement 3, "Criteria for Protective Action Recommendations for Severe Accidents," issued July 1996. Supplement 3 provides a protective action strategy development tool based on recent technical information and is intended for use by nuclear power reactor licensees to develop site-specific protective action recommendation procedures. Offsite response organizations should use Supplement 3 to develop protective action strategy guidance for decision makers. The guidance of Supplement 3 provides an acceptable method to comply with Appendix E to Part 50, Title 10 of the Code of Federal Regulations (10 CFR) Section IV, paragraph 3 in the use of evacuation time estimates in the formulation of protective action recommendations (PARs) for the plume exposure emergency planning zone, and provides guidance for the provisions of 10 CFR 50.47(b)(10) in the development of a range of PARs. Supplement 3 also provides guidance to support the information in NUREG-0654/FEMA-REP-1 that the U.S. Nuclear Regulatory Commission finds to be an acceptable method of meeting the requirements in 10 CFR 50.47(b)(7) for the development of a public information program. However, licensees may identify alternative methods of compliance with these requirements.

The U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors (PWRs)," on September 13, 2004, as the primary vehicle for addressing and resolving concerns associated with Generic Safety Issue 191, "Assessment of Debris Accumulation on PWR Sump Performance." Additionally, the NRC staff developed a safety evaluation of industry-developed guidance to provide an accepted method for evaluating PWR sump performance as requested in Generic Letter 2004-02. However, the safety evaluation and the industry guidance document provide little guidance for assessing chemical effects. The licensees are to address chemical effects on a plant-specific basis.

This guide was developed by the United States Department of Transportation and Nuclear Regulatory Commission, and the Canadian Nuclear Safety Commission to provide direction on the standard format and content for applications for approval of packages used to transport Type B(U) and fissile materials. The guide is designed to facilitate the United States validation of Competent Authority approvals for export/import purposes and limit redundant technical reviews. A companion guide document has been developed and published in Canada as RD-364, "Joint Canada - United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages."

This safety evaluation report (SER) documents the technical review of the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (PVNGS), license renewal application (LRA) by the U.S. Nuclear Regulatory Commission (NRC) staff (the staff). By letter dated December 11, 2008, and supplemented by letter dated April 14, 2009, Arizona Public Service Company (APS) (the applicant) submitted the LRA in accordance with Title 10 of the Code of Federal Regulations, Part 54 "Requirements for Renewal of Operating Licenses for Nuclear Power Plants." APS requests renewal of the PVNGS operating licenses (facility operating license numbers NPF-41, NPF-51, and NPF-74) for a period of 20 years beyond the current expiration dates of midnight on June 1, 2025, for Unit 1; April 24, 2026, for Unit 2; and November 25, 2027, for Unit 3.

The U.S Nuclear Regulatory Commission (NRC) issues licenses for the possession and use of source material provided that proposed facilities meet NRC regulatory requirements and would be operated in a manner that is protective of public health and safety and the environment. Under the NRC environmental protection regulations in the Code of Federal Regulations (CFR), Title 10, Part 51, which implement the National Environmental Policy Act (NEPA) of 1969, issuance of a license to possess and use source material for uranium milling requires an environmental impact statement (EIS) or a supplement to an EIS.

This safety evaluation report documents the technical review of the Oyster Creek Generating Station license renewal application by the staff of the United States Nuclear Regulatory Commission.

This safety evaluation report documents the technical review of the Vermont Yankee Nuclear Power Station license renewal application by the United States Nuclear Regulatory Commission staff.

The objective of this report is to describe the regulatory processes that assures that any plant-specific licensing bases will provide reasonable assurance that the operation of nuclear power plants will not be inimical to the public health and safety to the end of the renewal period. It is on the adequacy of this process that the Commission has determined that a formal renewal licensing review against the full range of current safety requirements would not add significantly to safety and is not needed to assure that continued operation throughout the renewal term is not inimical to the public health and safety or common defense and security.

In this NUREG, the staff of the U.S. Nuclear Regulatory Commission (NRC) discusses the applicable regulations for the inservice testing of pumps and valves and the examination and testing of dynamic restraints (snubbers) at commercial nuclear power plants. The information in NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants," Revision 0, issued April 1995, and Revision 1, issued January 2005, has described this topic in the past. This NUREG report replaces Revision 0 and Revision 1 to NUREG-1482 and is applicable, unless stated otherwise, to all editions and addenda to the American Society of Mechanical Engineers Code for Operation and Maintenance of Nuclear Power Plants (OM Code), which Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(b) incorporates by reference. In addition, the staff discusses other inservice test program topics such as the NRC process for the review of the OM Code, conditions on the use of the OM Code, and interpretations of the OM Code.

This Standard Review Plan is intended for use by the U.S. Nuclear Regulatory Commission (NRC) reviewer. It provides guidance for the safety review of license (specific or general) and certificate of compliance (CoC) renewal applications submitted by licensees and holders of CoCs for dry cask storage systems (DCSSs), respectively, as codified in Title 10 of the Code of Federal Regulations (10 CFR) Part 72, "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste." A license authorizes a licensee to store spent fuel in an NRC-approved DCSS at a site under the requirements of 10 CFR Part 72. To renew a specific license, an applicant must submit a license renewal application at least 2 years before the expiration of the license in accordance with the requirements of 10 CFR 72.42(b). To renew a general license, the general licensee or the CoC holder must submit a renewal application at least 30 days before the expiration of the associated CoC in accordance with the requirements of 10 CFR 72.240(b). The NRC may renew a specific license or a general license for a term not to exceed 40 years, in accordance with 10 CFR 72.42(a) or 10 CFR 72.212(a)(3), respectively.

This safety evaluation report documents the technical review of the Beaver Valley Power Station, Units 1 and 2, license renewal application by the United States Nuclear Regulatory Commission staff.

Computational fluid dynamics is used to predict the natural circulation flows between a simplified reactor vessel and the steam generator of a pressurized-water reactor (PWR) during a severe accident scenario. The results extend earlier predictions of steam generator inlet plenum mixing with the inclusion of the entire natural circulation loop between the reactor vessel upper plenum and the steam generator. Tube leakage and mass flow into the pressurizer surge line are also considered. The predictions are utilized as a numerical experiment to improve the basis for simplified models applied in one-dimensional system codes that are used during the prediction of severe accident natural circulation flows. An updated inlet plenum mixing model is proposed that accounts for mixing in the hot leg too. The new model is consistent with the predicted behavior and accounts for flow into a side mounted surge line if present. A density- based Froude number correlation is utilized to provide a method for determining the flow rate from the vessel to the hot leg directly from the conditions at the ends of the hot leg pipe. This provides a physically based approach for establishing the hot leg flows. The mixing parameters and correlations are proposed as a best-estimate approach for estimating the flow rates and mixing in one-dimensional system codes applied to severe accident natural circulation conditions. Sensitivity studies demonstrate the applicability of the approach over a range of conditions. The predictions are most sensitive to changes in the steam generator secondary side temperatures or heat transfer rates to the steam generator. Grid independence is demonstrated through comparisons with previous models and by increasing the number of cells in the model. A further modeling improvement is suggested regarding the application of thermal entrance effects in the hot leg and surge line. This work supports the U.S. Nuclear Regulatory Commission studies of steam generator tube integrity under severe accident conditions.

The United States Nuclear Regulatory Commission has updated NUREG-1650, Rev. 1, "The - United States Third National Report for the Convention on Nuclear Safety," issued September 2004, and will submit this report for peer review at the fourth review meeting of the Convention on Nuclear Safety at the International Atomic Energy Agency in April 2008. This report addresses the safety of land-based commercial nuclear power plants in the U.S. It demonstrates how the U.S. Government achieves and maintains a high level of nuclear safety worldwide by enhancing national measures and international cooperation, and by meeting the obligations of all the articles established by the Convention. These articles address the safety of existing nuclear installations, the legislative and regulatory framework, the regulatory body, responsibility of the licensee, priority to safety, financial and human resources, human factors, quality assurance, assessment and verification of safety, radiation protection, emergency preparedness, siting, design and construction, and operation. New to this update is a contribution by the Institute of Nuclear Power Operations (INPO) describing work done by the U.S. nuclear industry to ensure safety. Since the prime responsibility for the safety of a nuclear installation rests with the license holder, it was decided to involve the nuclear industry collectively through INPO, in explaining how the nuclear industry maintains and improves nuclear safety.

The Department of Homeland Security's Performance and Accountability Reports provide information that enables the President, Congress and the public to assess the effectiveness of the Department's mission performance and stewardship of resources. In participation with the Office of Management and Budget (OMB) alternative approaches to Performance and Accountability Reporting, the Department publishes the reports listed here rather than the consolidated Performance and Accountability Report published in previous years.

The Multi-Agency Radiation Survey and Assessment of Materials and Equipment manual (MARSAME) is a supplement to the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) providing information on planning, conducting, evaluating, and documenting radiological disposition surveys for the assessment of materials and equipment. MARSAME is a multi-agency consensus document that was developed collaboratively by four Federal agencies having authority and control over radioactive materials: Department of Defense (DOD), Department of Energy (DOE), Environmental Protection Agency (EPA), and Nuclear Regulatory Commission (NRC). The objective of MARSAME is to provide a multi-agency approach for planning, performing, and assessing disposition surveys of materials and equipment, while at the same time encouraging an effective use of resources.

As part of fire protection defense in depth, nuclear power plants and divided into separate fire areas by fire-rated structural barriers. Fire-rated penetration seals are installed to seal certain openings in these barriers. The seals maintain the fire-resistive integrity of the barriers and provide reasonable assurance that a fire will be confined to the area in which it started. The staff of the Fire Protection Engineering Section, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission, conducted a comprehensive technical assessment of penetration seals to address reports of potential problems, to determine if there were any problems of safety significance, and to determine if NRC requirements, review guidance, and inspection procedures are adequate. The staff did not find plant specific problems of the safety significance or concerns with generic implications. The staff concluded that the general condition of penetration seal programs in industry is satisfactory. The staff also concluded that actions it had taken in 1988 and 1994 to address potential penetration seal problems increased industry awareness of such problems and resulted in more thorough surveillances, maintenance, and corrective actions. These previous staff actions, together with continued licensee upkeep of existing penetration seal programs and continued NRC inspections, are adequate to maintain public health and safety. The staff recommended several minor revisions to the NRC fire protection regulation and review guidance.

The safety evaluation report documents the technical review of the Vogtle Electric Generating Plant, Units 1 and 2 license renewal application by the United States Nuclear Regulatory Commission staff.

The Standard Review Plan for Spent Fuel Dry Storage Facilities (FSRP) provides guidance to the staff of the U.S. Nuclear Regulatory Commission for reviewing applications for license approval or renewal for commercial independent spent fuel storage installations (ISFSIs). An ISFSI may be co-located with a reactor or may be away from a reactor site. These installations may be designed for the storage of irradiated nuclear fuel and associated radioactive materials.

As part of its redesign of the materials license program, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Material Safety and Safeguards (NMSS) has consolidated and updated numerous decommissioning guidance documents into a three-volume NUREG.