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The U.S. Nuclear Regulatory Commission (NRC) is addressing the human performance aspects of changes to operator actions that are credited for safety, especially those involving changes in the licensing basis of the plant, e.g., using a manual action in place of an automatic action for safety system operations. This document provides guidance for reviewing changes to human actions that are credited for safety. In this document, the term human action and operator action are used synonymously because most of the types of actions discussed are performed by operations staff. The evaluation method uses a two-phase approach. The first phase is a screening analysis of the licensee's proposed modification and the affected human actions to assess their risk importance. A graded, risk-informed approach is used to determine the appropriate level of human factors engineering review. This approach can be accomplished for submittals by licensees that are either risk-informed or non-risk-informed. For a risk-informed submittal, the first phase has four steps: Use of Regulatory Guide 1.174 to determine the risk importance of the entire plant change or modification which involves the human action; a quantification of the risk importance of the human action itself; a qualitative evaluation of the human action; and, an integrated assessment to determine the appropriate level of human factors engineering review. The proposed human actions are assigned to one of three risk levels (high, medium, low) as a result of phase 1. The level of the human factors review in the second phase corresponds to these risk levels. In the second phase, human actions are reviewed using standard criteria in human factors engineering to verify that the proposed action can be reliably performed when required. Human actions in the high risk level receive a detailed human factors engineering review and those in the medium risk level undergo a less detailed one, commensurate with their risk. For human actions placed in the low risk level, there is a minimal human factors review or none. The NRC review of licensee submittals that involve changes to human actions is an iterative process. The final results of the human factors review will provide input to the Integrated Decision- making and to a Safety Evaluation Report.

The intent of this technical report is to provide guidance on, and to assist applicants and licensees in, the implementation of Title 10 of the Code of Federal Regulations (10 CFR) Part 37, "Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material." This document describes methods that the U.S. Nuclear Regulatory Commission (NRC) finds acceptable for implementing the regulations.

Decommissioning means permanently removing a nuclear facility from service and reducing radioactive material on the licensed site to levels that permit termination of the NRC license.

Section 208 of the Energy Reorganization Act of 1974, as amended (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.

This report provides the Commission with an overview of the U.S. Nuclear Regulatory Commission (NRC) Office of Investigations' (OI) activities, mission and purpose, along with the framework of case inventory with highlights of significant cases that the NRC OI completed during Fiscal Year 2012 (reference SRM COMJC-89-8, dated June 30, 1989). This is the 24th OI Annual Report.taff to proceed with the development and implementation of the incentives for licensees to maintain an effective CAP.

The U.S. Nuclear Regulatory Commission (NRC) has developed quantitative methods, known as "Fire Dynamics Tools" (FDTs), for analyzing the impact of fire and fire protection systems in nuclear power plants (NPPs). These methods have been implemented in spreadsheets and taught at the NRC's quarterly regional inspector workshops. The FDTs were developed using state-of-the-art fire dynamics equations and correlations that were preprogrammed and locked into Microsoft Excel(R) spreadsheets. These FDTs enable inspectors to perform quick, easy, first- order calculations for potential fire scenarios using today's state-of-the-art principles of fire dynamics. Each FDTs spreadsheet also contains a list of the physical and thermal properties of the materials commonly encountered in NPPs.

The U.S. Nuclear Regulatory Commission has updated the U.S. National Report for the Convention on Nuclear Safety. This report will be submitted for peer review at the second review meeting of the Convention on Nuclear Safety at the International Atomic Energy Agency in April 2002. The scope of the report is limited to the safety of land-based civil nuclear power plants. The report demonstrates how the U.S. Government meets the main objective of the Convention -- to achieve and maintain a high level of nuclear safety worldwide. The report also shows that the U.S. Government meets the obligations of each of the articles established by the Convention. Specifically, 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, construction, and operation.

This report provides a comprehensive overview of the U.S. Nuclear Regulatory Commission's (NRC's) decommissioning program. Its purpose is to provide a stand-alone reference document which describes the decommissioning process and summarizes the current status of all decommissioning activities including the decommissioning of complex decommissioning sites, commercial reactors, research and test reactors, uranium mill tailings facilities, and fuel cycle facilities. In addition, this report discusses accomplishments in the decommissioning program since publication of the 2003 annual report (SECY-03-0161), and it identifies the key decommissioning program issues which the staff will address in fiscal year (FY) 2005.

On December 18, 2008, the U.S. Nuclear Regulatory Commission (NRC) staff provided a Policy Issue Notation Vote Paper, SECY-08-0197 (ADAMS Accession No. ML083360582), to the Commission which presented the regulatory and technical options of moving, or not moving, towards a greater degree of alignment of the NRC radiation protection regulatory framework with the International Commission on Radiological Protection (ICRP) Publication 103. In a Staff Requirements Memorandum dated April 2, 2009, SRM-SECY-08-0197 (ML090920103), the Commission approved the staff's recommendation to immediately begin engagement with stakeholders and interested parties to initiate development of the technical basis for possible revision of the NRC's radiation protection regulations, as appropriate and where scientifically justified. As part of the outreach to stakeholders and interested parties, NRC staff noted the need to expand the current occupational radiation dose information contained in the Radiation Exposure Information and Reporting System (REIRS) database.e improved to support current requirements for risk-informed, performance-based (RI/PB) applications.

This report is not a substitute for U.S. Government regulations, and compliance with the information and guidance provided is not required. The technical approaches, software, and methods described in these conference proceedings are provided for information only. Publication of these proceedings does not necessarily constitute Federal agency approval or agreement with the information contained herein. Use of product or trade names is for identification purposes only and does not constitute endorsement or recommendation for use by any Federal agency.

This is the U.S. Nuclear Regulatory Commission (NRC) staff's report of its monitoring of U.S. Department of Energy (DOE) non-high-level waste disposal actions in Calendar Year 2011, in accordance with Section 3116(b) of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (the NDAA). Section 3116 of the NDAA requires that (1) DOE consult with the NRC on its non-high-level waste determinations and plans, and (2) the NRC, in coordination with the covered States of South Carolina and Idaho, monitor disposal actions that DOE takes to assess compliance with NRC regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 61, "Licensing Requirements for Land Disposal of Radioactive Waste," Subpart C, "Performance Objectives." The NRC has prepared this report in accordance with NUREG-1854, "NRC Staff Guidance for Activities Related to U.S. Department of Energy Waste Determinations," dated August 2007 (NRC, 2007a).

The NRC regulates the operation of the civilian nuclear power plant fleet by establishing and enforcing regulatory requirements for their design, construction, and operation. To protect the health and safety of the public and environment, the NRC requires all nuclear power plants to have a spent fuel pool where the used reactor fuel assemblies are allowed to cool for a number of years before being moved to interim or permanent storage. Spent fuel pools (SFP) are robust structures with an extremely low likelihood of a complete loss of coolant under traditional accident scenarios. However, in the wake of the terrorist attacks of September 11, 2001, the SFP accident progression was reconsidered and reevaluated using best-estimate accident codes.

The U.S. Nuclear Regulatory Commission prepared this supplemental final environmental statement in response to the Tennessee Valley Authority application for facility operating license.

Boiling water reactor plants are equipped with safety/relief valves to protect the reactor from over-pressurization. Plant operational transients, such as turbine trips, will activate the SRV.

This report documents a comprehensive comparison of cross sections calculated using different methodologies and codes, including CASMO, HELIOS, and TRITON XS. The conclusions from this study have resulted in this guidance document on how to choose cross section histories and branches for boiling water reactor (BWR) analysis, and the methodology to collapse the fine-energy and -space fluxes calculated by the detailed lattice calculation. The guidance herein is applicable to all BWR designs.

The purpose of this document is to support the technical basis for the choice of test methods to assess fuel-rod cladding behavior following a loss-of-coolant accident (LOCA). For non-deformed cladding regions with uniform levels of hydrogen content and oxidation, and for irradiated cladding, the ring compression test (RCT) has been selected. The four-point bend test (4-PBT) has been selected for evaluating the performance of ballooned and ruptured regions of a fuel rod in LOCA analysis. The technical basis is founded on the results of the NRC's LOCA research program, which was designed to measure the mechanical behavior of both non-deformed and ballooned and ruptured cladding following LOCA conditions. Integral LOCA tests were conducted at Argonne National Laboratory and Studsvik Laboratory in Sweden. The Japanese Atomic Energy Agency (JAEA) has also performed LOCA integral experiments. The results and observations of these experimental programs have been combined to develop considerations regarding the impact of oxidation and hydrogen content on the mechanical behavior of ballooned and ruptured cladding following LOCA conditions.

In its Reactor Oversight Process (ROP), the U.S. Nuclear Regulatory Commission (NRC) currently uses performance indicators to quantify safety system unavailability (SSU) for four important nuclear power plant systems. Over time, the NRC staff has identified a number of concerns related to the use of these indicators, including the use of short-term unavailability to approximate unreliability, the use of generic performance thresholds irrespective of variations in risk significance, and potential double-counting as a result of support system failures cascading onto front line systems. Moreover, the way the SSU indicators currently measure unavailability is inconsistent with the definition in the NRC's Maintenance Rule, as well as the indicators promulgated by the World Association of Nuclear Operators and the Institute of Nuclear Power Operations.

In the staff requirements memorandum (SRM) for SECY-10-0031, "Revising the Fuel Cycle Oversight Process," dated August 4, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102170054), the Commission directed the U.S. Nuclear Regulatory Commission (NRC) staff to consider how the NRC Enforcement Policy could best reflect that most fuel cycle licensees have voluntarily developed corrective action programs (CAPs). In response to the Commission's direction, the staff revised the NRC Enforcement Policy to disposition Severity Level IV violations as noncited violations if the NRC determines that the licensee's CAP is effective, the licensee enters the violation in its CAP, and other criteria are met, as delineated in Section 2.3.2 of the NRC Enforcement Policy (ADAMS Accession No. ML12340A295). In SRM-SECY-11-0140, "Enhancements to the Fuel Cycle Oversight Process," dated January 5, 2012 (ADAMS Accession No. ML120050322), the Commission directed staff to proceed with the development and implementation of the incentives for licensees to maintain an effective CAP.

This report is part of a series of updates of technical issues concerning fitness for duty in the nuclear power industry. It discusses technologies relevant to the detection and management of two key elements of a fitness-for-duty program: drug and alcohol testing and fatigue management. On drug and alcohol testing, the report provides an introduction to the pharmacokinetics of drugs of abuse in different bodily fluids and substances (matrices), a review of the technologies used to separate, identify, and quantify drugs in workplace drug testing programs, and a description of emerging research in developing and validating the technology systems capable of testing alternative matrices as well as newly appearing drugs of abuse, both in the laboratory and at the point of collection. On fatigue management, the report reviews recent research on sleep and fatigue, describes efforts under way to develop and deploy technologies to aid fatigue assessment and management, reviews the status of fatigue management in industries and governmental sectors where fatigue is a significant safety concern, and discusses implications for the nuclear power industry. Finally each chapter includes an extensive bibliography of documents to support further, more in-depth reviews.

The U.S. Department of Energy (DOE) has contracted with Duke Cogema Stone & Webster (DCS) to design, construct, and operate a proposed Mixed Oxide (MOX) Fuel Fabrication Facility that would convert depleted uranium and weapons-grade plutonium into MOX fuel. The proposed MOX facility would be located on the DOE's Savannah River Site in South Carolina. Use of the proposed facility to produce MOX fuel would be part of the DOE's surplus plutonium disposition program. The purpose of the DOE program is to ensure that plutonium produced for nuclear weapons and declared excess to national security is converted to proliferation- resistant forms.

At the fifth review meeting of the Convention on Nuclear Safety (CNS), held April 4 - 15, 2011, in Vienna, Austria, the contracting parties agreed to analyze the relevant issues of the Fukushima accident during an extraordinary meeting. This meeting is scheduled to take place August 27 - 31, 2012, in Vienna, Austria. The objectives of this CNS extraordinary meeting are to enhance safety through reviewing and sharing lessons learned and actions taken by contracting parties in response to the events of Fukushima and to review the effectiveness and continued stability of the provisions of the CNS.

Each commercial nuclear power plant operating in the United States has a comprehensive Fire Protection Program (FPP) that has been reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC). 1 The purpose of the FPP is to prevent the occurrence of fire and minimize radioactive releases to the environment in the unlikely event a significant fire were to occur. To achieve these objectives, the FPP integrates a number of plant designs and operating features (i.e., systems, components, personnel and administrative controls) needed to provide defense-in-depth protection of the public health and safety. This means that the FPP includes measures that are directed at reducing the likelihood of fires and explosions; rapidly detecting and suppressing those fires that may occur, and ensuring the capability to achieve and maintain safe shutdown conditions in the event of a significant fire.

In an ongoing effort to increase effectiveness and efficiency through improved prioritization of regulatory activities, a decision process has been developed to aid in the determination of risk significance of Emergency Preparedness (EP) program elements. The DedUctive Quantification Index (DUQI) method was developed and used in a proof of concept application for two representative nuclear power plant sites. The results show the cumulative population dose is reduced through implementation of a formal EP program compared to conditions in which an emergency response would be implemented in an ad hoc manner. Dose was shown to be consistently lower for all analyses. The DUQI method was also applied to determine risk significance of specific EP elements. Analyses included a response where sirens are assumed not operable in the 2-5 mile area around the nuclear power plant, and for a delay in notification to offsite response organizations. Detailed consequence analysis modeling was performed using site specific information. The process used information from historical studies, such as NUREG-1150 combined with current knowledge. Data for specific sites was used in selected areas to increase the credibility of the product, but the results are not applicable to any specific site. Improvements were made to the modeling approach by simulating evacuee road loading in greater detail than previous studies. The 95th percentile cumulative population dose results were produced and used to support the study conclusions.

Under normal operating conditions, cladding and core structural materials operate around 300 degrees Celsius (C), and fuel pellets experience peak temperatures below 2,000 degrees C at the pellet centerline. At these temperatures, fuel cladding integrity is maintained, and fission products are contained within the fuel rods. However, under some abnormal conditions, higher temperatures and other conditions significantly alter the behavior of these materials. These conditions can threaten core coolability and lead to fission product release. This NUREG report considers the following two types of accident conditions: (1) reactivity-initiated accidents and (2) loss-of-coolant accidents. This report describes the fuel behavior of each accident condition from basic concepts to the current state of the art. It also mentions safety criteria, references the classic experimental work in each of these areas, and presents equations and figures that permit some quantitative evaluations.

This report has been prepared to support technical discussion of and planning for future research supporting implementation of burnup credit for boiling-water reactor (BWR) spent fuel storage in spent fuel pools and storage and transport cask applications. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries, including experience with burnup credit for pressurized-water reactor (PWR) spent fuel. Relevant physics and analysis phenomena are identified, and an assessment of their importance to burnup credit implementation is given. Results from sensitivity studies of some of the key phenomena are presented.

Two of the many elements contributing to the safety of nuclear power are emergency response and the feedback from operating experience into plant operations. These are achieved partly by the licensee event reporting requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.72, "Immediate Notification Requirements for Operating Nuclear Power Reactors," and 10 CFR 50.73, "Licensee Event Report System." In 10 CFR 50.72, the U.S. Nuclear Regulatory Commission (NRC) provides for immediate notification requirements through the emergency notification system, and in 10 CFR 50.73 provides for 60-day written licensee event reports.

Under the provisions of the Atomic Energy Act and Pursuant to Title 10 of the U.S. Code of Federal Regulations Part 40, the U.S. Nuclear Regulatory Commission is considering whether to issue a license that would allow international Isotopes Fluorine Products, Incorporate to possess, use, transfer, or deliver source and byproduct materials at a proposed fluorine extraction and depleted uranium deconversion facility near Hobbs in Lea County, New Mexico.

This report proposes and documents a computational benchmark for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup- credit cask designed to hold 68 boiling water reactor (BWR) spent nuclear fuel assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin for partial burnup credit (ISG-8) and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. This benchmark model will also be used as the base case in future sensitivity studies, to be documented in a companion NUREG/CR. The geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission- product margin. The reference solutions were generated with the SCALE 6.1 package. The SCALE depletion and criticality sequences have been extensively validated elsewhere for a variety of applications, including light water reactor fuel. Note that the reference solutions presented in this report are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1, 8.24, and 8.27 requirements for validation of calculational methods and is not intended to be used to establish biases and bias uncertainties for burnup-credit analyses.

This supplement to the final supplemental environmental impact statement (FSEIS) for the proposed license renewal of Indian Point Nuclear Generating Unit Nos. 2 and 3 incorporates new information that the U.S. Nuclear Regulatory Commission (NRC) staff has obtained since the publication of the FSEIS in December 2010.

This report contains the information presented at the meeting of the Working Group on Performance Confirmation Plans for the Proposed Yucca Mountain High-Level-Waste Repository.