Bøger af U S Nuclear Regulatory Commission

RASCAL 4 consists of several modules. Consequence assessments for nuclear power plants use five of the modules. The code invokes four of these modules when the user selects "Source Term to Dose" on the opening screen. The first module calculates the time-dependent atmospheric release source term. The atmospheric release source term is the rate at which radioactive material is released to the environment. It also includes other information that defines how the release takes place. The second and third modules perform the atmospheric transport, dispersion, and deposition calculations and the dose calculations. The fourth module is used to create the meteorological data file used by the atmospheric transport, dispersion, and deposition modules. The fifth module is used for intermediate-phase dose calculations based on field measurements. Uranium fuel cycle consequence assessments use the sixth module.

To the knowledge of the authors, these studies are the first hydraulic characterizations of a full length, highly prototypic 17×17 pressurized water reactor (PWR) fuel assembly in low Reynolds number flows. The advantages of full scale testing of prototypic components are twofold. First, the use of actual hardware and dimensionally accurate geometries eliminates any issues arising from scaling arguments. Second, many of the prototypic components contain intricacies by design that would not be reproduced by using simplified flow elements. While this approach yields results that are inherently specific to the fuel assembly under testing, the differences in commercial designs are considered minor, particularly when considering the hydraulics of the entire assembly.

Various forms of degradation have been observed in the containment vessels of a number of operating nuclear power plants in the United States. Examples of degradation include corrosion of the steel shell or liner, corrosion of reinforcing bars and prestressing tendons, loss of prestressing, and corrosion of bellows. The containment serves as the ultimate barrier against the release of radioactive material into the environment. Because of this role, compromising the containment could increase the risk of a large release in the unlikely event of an accident. Previous work in this area has assessed the effects that degradation has on the pressure retaining capacity of the containment vessel through structural analysis that account for degradation. These analyses have provided useful information about the effects of the degradation on the structural capacity of the containment in both deterministic and probabilistic fashions. However, the appropriate metric to use in assessing containment degradation effects during a severe accident was determined to require additional study. The previous work with probabilistic descriptions of the containment capacity were obtained from the results of the structural analysis models, and used as input for the risk models (Probabilistic Risk Assessment analyses). The risk was formulated in terms of the large early release frequency (LERF). The relative LERF values were computed for various postulated cases of degradation. In that study, an instance of degradation was treated as a change in the plant's licensing basis and assessed with U.S. NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." The Regulatory Guide provides the limits of the acceptable increases in LERF due to changes in the plant. Many of the cases of postulated degradation were those consisting of local corrosion in the liner or shell that produce leaks that do not contribute significantly to LERF, and in some cases, cause no change in LERF. Early releases due to small leaks were found to contribute to the small early release frequency (SERF). Since Regulatory Guide 1.174 does not provide guidance on the limits of SERF, additional deterministic analyses were performed in this study to assess the effects of degradation on the consequences using metrics other than LERF. This study was performed using the Sandia codes MELCOR and MACCS. These codes were used to simulate two different accident scenarios (long- and short-term station black-outs) and compute the resulting consequences for a PWR plant with a reinforced concrete containment. The structural analyses used in the previous probabilistic study were used to develop the containment behavior models for the accident simulations. Several different postulated cases of liner corrosion were considered to enable a comparison of the consequences.

In 1962, the Atomic Energy Commission issued Technical Information Document 14844, "Calculation of Distance Factor for Power and Test Reactors". In this document, a release of fission from the core of a light water reactor into the containment atmosphere was postulated for the purpose of calculating off-site doses in accordance with 10 CFR part 100.

Cyber-Physical Systems (CPS) are hybrid networked cyber and engineered physical elements co- designed to create adaptive and predictive systems for enhanced performance.1, 2 These smart systems present a key opportunity to create a competitive advantage for U.S. industrial innova- tion and to improve the performance and reliability of new and existing systems. From smart manufacturing and the electric smart grid, to smart structures and transportation systems, CPS will pervasively impact the economy and society.

The staff of the U.S. Nuclear Regulatory Commission (NRC) is issuing Revision 2 to the NUREG-1482, "Guidelines for Inservice Testing at Nuclear Power Plants," to assist the nuclear industry in establishing a basic understanding of the regulatory basis for pump and valve inservice testing (IST) programs and dynamic restraint (snubbers) examination and testing programs. This NUREG also provides information regarding the NRC's involvement in the development of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code). In this NUREG, the staff discusses OM Code inquiries, the inservice examination and testing of snubbers, pump and valve IST, the use of ASME code cases, conditions on the use of the OM Code, guidance for OM Code noncompliance, requests for alternatives to the OM Code at operating commercial nuclear power plants, and the development of IST programs for new reactors*.

Research is being conducted for the U.S. Nuclear Regulatory Commission (NRC) at the Pacific Northwest National Laboratory (PNNL) to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the detection and characterization of flaws in nuclear power plant components. One area of concern relates to the nickel-based alloys used in primary pressure boundary components in pressurized water reactors (PWRs). Nickel-based alloys exposed to reactor coolant in PWRs may experience a form of degradation known as primary water stress corrosion cracking (PWSCC). One PWR component that has an operational history of PWSCC is the control rod drive mechanism (CRDM) nozzle. The CRDM nozzles are cylindrical penetrations in the upper reactor pressure vessel (RPV) head that allow for the insertion and removal of control rods. The penetration tube is held in place with an interference fit, and is seal-welded on the underside of the vessel head with a J-groove weld. Cracking in the nozzle or weld metal can allow borated water to leak to the top of the RPV head. Boric acid corrosion of the RPV head, as occurred at Davis Besse, is a concern, as is nozzle ejection in the presence of extensive circumferentially oriented cracking. In response to a number of observations of RPV head leakage at domestic plants, NRC regulations were modified to require PWR licensees to periodically perform a demonstrated surface or volumetric leak path assessment of all J-groove welds in the RPV head.

The Standard Review Plan For Dry Cask Storage Systems provides guidance to the Nuclear Regulatory Commission staff in the Spent Fuel Project Office for performing safety reviews of dry cask storage systems.

This report provides a comprehensive overview of the Decommissioning Program of the U.S. Nuclear Regulatory Commission. Its purpose is to provide a stand-alone reference document that describes the decommissioning process and summarizes the status of decommissioning activities, under NRC and Agreement State jurisdiction, from October 1, 2007 through September 30, 2008.

Under the auspices of an addendum to the memorandum of understanding between the Electric Power Research Institute and the U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research for cooperative research, a pilot study has been completed to evaluate the feasibility of developing a fully probabilistic, fracture-mechanics-based computational tool to evaluate the rupture probability of reactor coolant piping. This project, known as xLPR for Extremely Low Probability of Rupture, is initially focused on evaluating pipe rupture probabilities within Alloy 82/182 dissimilar metal welds located in lines licensed for leak-before-break (LBB) as allowed under General Design Criterion 4, "Environmental and Dynamic Effects Design Bases," of Appendix A, "General Design Criteria for Nuclear Power Plants," to Title 10 of the Code of Federal Regulations Part 50, "Domestic Licensing of Production and Utilization Facilities." The current LBB regulatory basis does not allow for assessment of piping systems subject to active degradation mechanisms, such as primary water stress-corrosion cracking, which has been detected in some systems that have been granted LBB approval. Although the piping systems susceptible to this type of corrosion have been shown through deterministic arguments to comply with the regulations, no fully probabilistic tool currently exists to directly assess this compliance.

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.

This report describes the events that occurred on a drill rig Montana on May 21, 2002, that led to the unplanned radiation exposure of 31 rig workers.

The U.S. Nuclear Regulatory Commission (NRC) has prepared Revision 3 to NUREG-1650, AUnited States of America Fifth National Report for the Convention on Nuclear Safety@ for submission for peer review at the fifth review meeting of the Convention on Nuclear Safety, to be convened at the International Atomic Energy Agency in Vienna, Austria, in April 2011. The NRC issued the fourth report in September 2007. This revised 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, the priority given 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.

Section 208 of the Energy Reorganization Act of 1974 (Public Law 93-438) defines an "abnormal occurrence" (AO) as an unscheduled incident or event that the U.S. Nuclear Regulatory Commission (NRC) determines to be significant from the standpoint of public health or safety. The Federal Reports Elimination and Sunset Act of 1995 (Public Law 104-66) requires that the NRC report AOs to Congress annually.

As a result of a major fire that occurred at the Browns Ferry Nuclear Power Plant in 1975, the U.S. Nuclear Regulatory Commission (NRC) significantly revised its regulatory framework to enhance fire protection programs (FPPs) at operating nuclear power plants (NPPs). The revised criteria used in this framework had three main objectives to (1) prevent significant fires, (2) ensure the capability to shut down the reactor and maintain it in a safe-shutdown condition, and (3) minimize radioactive releases to the environment in the event of a significant fire.

This framework document will be used as the guiding document for the creation of construction inspection manual chapters and inspection procedures to support the 10 CFR Part 52 licensing process.

The U.S. Nuclear Regulatory Commission (NRC) encourages design certification and combined license applicants to use this guidance to optimize physical security during the design phase. The expected result is a more robust security posture with less reliance on operational programs (human actions) and potentially costly retrofits. The NRC also encourages operating reactor licensees to use this guidance in planning and executing changes and upgrades of physical protection systems at existing sites.

The United States and the Commission of European Communities conducted a series of expert elicitations to obtain distributions for uncertain variables used in health consequence analyses related to accidental release of nuclear material. The distributions reflect degrees of belief for non-site-specific parameters that are uncertain and are likely to have significant or moderate influence on the results. The present work presents the effort to develop ranges of values and degrees of belief that fairly represent the divergent opinions of the experts while maintaining the resulting parameters within physical limits. Where necessary, there is a discussion of correlation coefficients that should be included when the uncertainty is used in a calculation. The methodology used a resampling of the experts' values and was based on the assumption of equal weights of the experts' opinions. Various statistical properties of the distributions, the median, the mean, and the mode, are presented so that the user can choose a parameter value when only a point estimate is desired.

The "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants" (SRP-LR) provides guidance to Nuclear Regulatory Commission (NRC) staff reviewers in the Office of Nuclear Reactor Regulation (NRR). These reviewers perform safety reviews of applications to renew nuclear power plant licenses in accordance with Title 10 of the Code of Federal Regulations (CFR) Part 54. The principal purposes of the SRP-LR are to ensure the quality and uniformity of staff reviews and to present a well-defined base from which to evaluate applicant programs and activities for the period of extended operation. The SRP-LR also is intended to make regulatory information widely available to enhance communication with interested members of the public and the nuclear power industry and to improve their understanding of the staff review process.

This report provides guidance to applicants that produce radioactive materials using an accelerator. It provides guidance to an applicant in preparing a license application as well as NRC criteria for evaluating the license application.

This document is a safety evaluation report (SER) on the license renewal application (LRA) for Columbia Generating Station (Columbia), as filed by Energy Northwest (the applicant). By letter dated January 19, 2010, Energy Northwest submitted its application to the U.S. Nuclear Regulatory Commission (NRC) for renewal of the Columbia operating license for an additional 20 years. The NRC staff (the staff) prepared this report to summarize the results of its safety review of the LRA for compliance with Title 10, Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants," of the Code of Federal Regulations (10 CFR Part 54). The NRC project manager for the license renewal review is Arthur D. Cunanan.

The ultimate heat sink is defined as the complex of cooling-water sources necessary to safely shut down and cool down a nuclear power plant. Cooling ponds, spray ponds, and mechanical draft cooling towers are some examples of the types of ultimate heat sinks in use today.

This guidance document sets forth means, methods and procedures that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for satisfying the requirements for the physical protection of spent nuclear fuel (SNF) during transportation by road, rail and water, and for satisfying the requirements for background investigations of individuals granted unescorted access to SNF during transportation. Chapter 1 discusses the regulatory basis and definitions applicable to shipments of SNF. Chapter 2 corresponds to the general requirements for the physical protection of SNF while in transit. These requirements apply to all SNF shipments regardless of the mode of transportation used for a particular shipment. Chapters 3, 4, and 5 discuss additional requirements specific to each particular transport mode-road, rail, or U.S. waters. Chapter 6 discusses the requirements for background investigations of individuals with unescorted access to SNF in transit.

The nuclear industry and the U.S. Nuclear Regulatory Commission explicitly recognized the importance of management and organizational fact to nuclear facility safety in aftermath of the accident at Three Mile Island Unit 2.

The report summarizes the results of a 5-year study conducted by the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research. The aim of this study was to develop the technical basis for revision of the Pressurized Thermal Shock Rule.

This document supplements the draft supplemental environmental impact statement (DSEIS) which had been prepared in response to an application submitted by NextEra Energy Seabrook, LLC (NextEra) to renew the operating license for Seabrook Station (Seabrook) for an additional 20 years. This supplement incorporates new information that the U.S. Nuclear Regulatory Commission (NRC) staff has obtained since the publication of the DSEIS in August 2011.

This report was prepared for the Advisory Committee on Reactor Safeguards to provide a basis for discussing possible changes to the general design criteria of Appendix A to Tile 10, Part 50.

The backfitting process is the process by which the U.S. Nuclear Regulatory Commission decides whether to issue new or revised requirements or staff positions to licenses of nuclear power reactor facilities.

The ultimate heat sink is defined as the complex of sources service or house water supply necessary to safely operate, shut down, and cool down a nuclear power plant.

This report provides guidance to an applicant applying for a commercial radiopharmacy license, as well as providing NRC with the appropriate criteria for evaluating such applications.