Bøger af U S Nuclear Regulatory Commission

This draft supplemental environmental impact statement has been prepared in response to an application submitted by Exelon Generation Company, LLC (Exelon) to renew the operating license for Limerick Generating Station, Units 1 and 2 (LGS) for an additional 20 years.

Instances of stress corrosion cracking in reactor pressure vessel internal components have been found, especially in the boiling water reactor (BWR) core shroud. Results from in-service inspection are an important aspect of integrity evaluations. One of the major goals of the work described in this report is to quantify the crack detection reliability and sizing error of ultrasonic inspection methods for reactor internals. A mockup is described, along with its application to the assessment on nondestructive evaluation reliability for in-service inspection of reactor internals. The mockup of a BWR core shroud includes cracks in some of the 40 welded assemblies selected to represent field conditions. The selected material and geometry include most conditions and alloys used in the core shroud and its support structure. This report provides an overview of the work being performed and focuses on the parametric study results relating ultrasonic response and its variance to inspection effectiveness. Results of a blind test of an in-service inspection vendor's phased array technique are also provided in this report.

General Design Criterion 1 of Appendix A to Part 50 of Title 10 of the Code of Federal Regulations requires that structures, systems, and components important to safety be designed to withstand the effects of earthquakes without a loss of capability to preform their safety function.

This document provides guidance to U.S. Nuclear Regulatory Commission staff in implementing the provisions in Title 10 of the Code of Federal Regulations Part 51 (10 CFR Part 51), "Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions" when conducting an environmental review for the renewal of a nuclear power plant operating license(s). This standard review plan guides the staff in preparing a plant-specific supplemental environmental impact statement to NUREG-1437, Revision 1, Generic Environmental Impact Statement for License Renewal of Nuclear Plants. This document supplements NUREG-1555, Standard Review Plans for Environmental Reviews for Nuclear Power Plants, which provides guidance for the environmental reviews of construction permits, initial operating licenses, early site permits, and combined licenses for new nuclear power plants.

The U.S. Nuclear Regulatory Commission has conducted several risk assessments and other analysis to evaluation the safety of transportation of spent power reactor nuclear fuel during the past 35 years.

The U.S. Nuclear Regulatory Commission (NRC) issues licenses for the possession and use of source and byproduct materials provided that facilities meet NRC regulatory requirements and will 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 of 1969 (NEPA), issuance of a license to possess and use source and byproduct materials during uranium recovery and milling requires an environmental impact statement (EIS) or a supplement to an EIS (SEIS).

This report documents the U.S. Nuclear Regulatory Commission's (NRC's) staff review and safety and safeguards evaluation of International Isotopes Fluorine Products, Inc.=s (IIFP), application for a license to construct a Fluorine Extraction Process & Depleted Uranium Deconversion Plant (FEP/DUP) and possess and use source materials. INIS proposes that the FEP/DUP be located in Lea County, New Mexico, approximately 24 kilometers (15 miles) west of the city of Hobbs. The facility will possess natural and depleted uranium, and significant quantities of hazardous chemicals derived from licensed material, e.g., hydrogen fluoride.

This safety evaluation report (SER) documents the technical review of the R.E. Ginna Nuclear Power Plant (Ginna) license renewal application (LRA) by the U.S. Nuclear Regulatory Commission staff (staff). By letter dated July 30, 2002, Rochester Gas & Electric Corporation (RG&E or the applicant) submitted the LRA for Ginna in accordance with Title 10, Part 54, of the Code of Federal Regulations (10 CFR Part 54 or the Rule). Through that LRA, RG&E requested that the U.S. Nuclear Regulatory Commission (NRC) renew the operating license for Ginna (license number DPR-18) for a period of 20 years beyond the current expiration of midnight, September 18, 2009.

The report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission's Radiation Exposure Information and Report System database.

Water availability has been a long-standing issue in the history of the arid Southwest United States. Natural precipitation is generally insufficient for farming.

This document provides U.S. Nuclear Regulatory Commission staff guidance conducting activities related to waste determinations.

The U.S. Nuclear Regulatory Commission (NRC) regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 52 include requirements for licensing new nuclear power plants.3 These regulations include the NRC's requirements for early site permit (ESP), design certification, and combined license (COL) applications. The ESP process (10 CFR Part 52, Subpart A, "Early Site Permits") is intended to address and resolve siting-related issues. The design certification process (10 CFR Part 52, Subpart B, "Standard Design Certifications") provides a means for a vendor to obtain NRC certification of a particular reactor design. Finally, the COL process (10 CFR Part 52, Subpart C, "Combined Licenses") allows an applicant to seek authorization to construct and operate a new nuclear power plant. A COL may reference an ESP, a certified design, both, or neither. As part of demonstrating that all applicable NRC requirements are met, a COL applicant referencing an ESP or certified design must demonstrate compliance with any requirements not already resolved as part of the referenced ESP or design certification before the NRC issues that COL.

At the request of Chairman Gregory B. Jaczko, a task force headed by Commissioner George Apostolakis prepared this report. The task force's charter was to develop a strategic vision and options for adopting a more comprehensive, holistic, risk-informed, performance- based regulatory approach for reactors, materials, waste, fuel cycle, and transportation that would continue to ensure the safe and secure use of nuclear material. The proposed risk management regulatory framework builds upon well established practices, such as the NRC's defense-in-depth philosophy and its policies to incorporate risk-informed and performance- based approaches into the agency's regulation and oversight of byproduct, source, and special nuclear materials. Risk management is being adopted by many different organizations, including Federal agencies, and would seem to be the logical next step in the evolution of the NRC's regulatory programs. The report describes a proposed risk management regulatory approach that could be used to improve consistency among the NRC's various programs and discusses implementing such a framework for specific program areas.

To more effectively fulfill its statutory responsibilities for the pre-licensing and licensing actions for the proposed high-level waste repository at the Yucca Mountain site, the U.S. Nuclear Regulatory Commission has focused its repository program on resolving issues most significant to overall system performance.

This document provides guidance on how to treat uncertainties associated with probabilistic risk assessment (PRA) in risk-informed decisionmaking. The objectives of this guidance include fostering an understanding of the uncertainties associated with PRA and their impact on the results of PRA and providing a pragmatic approach to addressing these uncertainties in the context of the decisionmaking.

This technical report contains information intended to provide program-specific guidance and assist applicants and licensees in preparing applications for materials licenses for fixed gauges. In particular, it describes the types of information needed to complete U.S. Nuclear Regulatory Commission (NRC) Form 313, "Application for Materials License." This document describes both the methods acceptable to the NRC license reviewers in implementing the regulations and the techniques used by the reviewers in evaluating the application to determine if the proposed activities are acceptable for licensing purposes.

Section 208 of the Energy Reorganization Act of 1974 defines an abnormal occurrence as an unscheduled incident or event which the U.S. Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety.

The susceptibility of steam generator tubes to degradation is affected by various factors, including the steam generator design, the operating environment, and operating and residual stress.

ABSTRACT Dry storage cask designs for spent nuclear fuel are submitted to the U.S. Nuclear Regulatory Commission (NRC) for certification under Title 10 of the Code of Federal Regulations (10 CFR) Part 72, "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High- Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste." The NRC staff technical review of these designs is performed in accordance with 10 CFR Part 72 and the "Standard Review Plan (SRP) for Spent Fuel Dry Storage Systems at a General License Facility" (NUREG-1536, 2010). To ensure that the cask and fuel material temperatures of the dry cask storage system will remain within the allowable limits or criteria for normal, off-normal, and accident conditions, a thermal review is performed as part of the application's technical review. Recent applications increasingly have used thermal-hydraulic analyses and computational fluid dynamics (CFD) codes (e.g., FLUENT) to demonstrate the adequacy of the thermal design. Therefore, in cooperation with the Division of Spent Fuel Storage and Transportation of the Office of Nuclear Material Safety and Safeguards, the Office of Nuclear Regulatory Research developed this guide to provide practical advice for reviewing CFD methods used in vendor applications and for achieving high-quality CFD simulations of a dry cask. To assist in the analysis, the report includes procedures, analysis methods, and acceptable assumptions.

The U.S. Nuclear Regulatory Commission is establishing "good practices" for performing human reliability analyses (HRAs) and reviewing HRAs to assess the quality of those analyses. The good practices were developed as part of the NRC's activities to address quality issues related to probabilistic risk assessment (PRA) and, as such, support the implementation of Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," dated February 2004.

During plant operation, the walls of reactor pressure vessels are exposed to neutron radiation, resulting in localized embrittlement of the vessel steel and weld materials in the core area.

There is a movement to introduce risk-informed and performance-based (RI/PB) analyses into fire protection engineering practice, both domestically and worldwide. This movement exists in both the general fire protection and the nuclear power plant (NPP) fire protection communities. The U.S. Nuclear Regulatory Commission (NRC) has used risk-informed insights as a part of its regulatory decision making since the 1990s.

By letter dated August 10, 2009, Powertech (USA), Inc. (Powertech) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for a new source and byproduct material license for the Dewey-Burdock In-Situ Uranium Recovery Project, located in Fall River and Custer Counties, South Dakota. The applicant is proposing to recover uranium using the in-situ leach (ISL) [also known as in-situ recovery (ISR)] process. The proposed Dewey-Burdock ISR Project would include processing facilities and sequentially developed wellfields sited in two contiguous areas, the Burdock area and the Dewey area. Proposed facilities include a central processing plant in the Burdock area, a satellite facility in the Dewey area, wellfields, Class V deep injection wells and/or land application areas for disposal of liquid wastes, and the attendant infrastructure (e.g., pipelines and surface impoundments).

This Guide has been developed in parallel with the rule-making on 10 CFR Parts 30,31, and 32, "Requirements for Certain Generally Licensed Industrial Devices Containing Byproduct Material".

Section 208 of the Energy Reorganization Act of 1974 defines an "abnormal occurrence" as an unscheduled incident or event that the U.S. Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety.

This report documents the U.S. Nuclear Regulatory Commission staff review and safety and safeguards evaluation of the Louisiana Energy Services application for a licenses to construct a gas centrifuge uranium enrichment facility and possess and use special nuclear material, source material, and byproduct material in gas centrifuge uranium enrichment facility.

If the renewed license is issued, the appropriate energy-planning decisionmakers, along with STPNOC, will ultimately decide if the reactor units will continue to operate based on factors such as the need for power. If the operating licenses are not renewed, then the facility must be shut down on or before the expiration dates of the current operating licenses-August 20, 2027, and December 15, 2028.

This compilation of statutes and materials pertaining to nuclear regulatory legislation through the 112th Congress, 2nd Session, has been prepared by the Office of the General Counsel, U.S. Nuclear Regulatory Commission, with the assistance of staff, for use as an internal resource document. The compilation is not to be used as an authoritative citation in lieu of the primary legislative sources. Furthermore, while every effort has been made to ensure the completeness and accuracy of this material, neither the United States Government, the Nuclear Regulatory Commission, nor any of their employees makes any expressed or implied warranty or assumes liability for the accuracy or completeness of the material presented in this compilation.

A program plan for nuclear power plant fire protection research has been presented in this report. The principal objective of the program is to create a data base that would reduce the uncertainties in fire probabilistic risk assessment of plants. A three-pronged approach of characterization of potential fires, determination of the ensuing environment, and determination of failure thresholds of safetyrelated equipment in that environment is described. The techniques are to be applied to estimating the fire safety margin, available in a control room.

The principle purpose of this SRP is to ensure the quality and uniformity of reviews conducted by the staff of the NRC's Office of Nuclear Material Safety and Safeguards.