License Renewal Environmental Fatigue Screening Application

2014 ◽  
Author(s):  
Christopher T. Kupper ◽  
Mark A. Gray
Keyword(s):  
Nuclear Regulatory Commission ◽  
Plant Component ◽  
Screening Process ◽  
The Us ◽  
Piping Systems ◽  
Class 1 ◽  
Major Equipment ◽  
Regulatory Commission ◽  
License Renewal ◽  
Detailed Application

In NUREG-1801 (GALL) Revision 0 and Revision 1, the US Nuclear Regulatory Commission (NRC) defined the locations evaluated in NUREG/CR-6260 as a minimum acceptable set for evaluation of environmentally assisted fatigue (EAF), in addressing license renewal for nuclear plant components. Within GALL Revision 2, the NRC revised the expectation, so that plants also investigate the possibility of other locations being more limiting. To address GALL Revision 2 and NUREG-1800 Revision 2, an EAF screening methodology was developed that considers all Safety Class 1 reactor coolant pressure boundary components in major equipment and piping systems that are susceptible to EAF, including those locations listed in NUREG/CR-6260. While the overall screening process steps are similar to those published by EPRI, elements of the detailed application of some steps were performed using alternative techniques. The screening process utilized the comprehensive database of plant component fatigue qualifications available in NSSS vendor documentation, and yielded a comprehensive list of lead indicator locations for EAF consideration. This paper describes the overall process and alternate methods in the context of a specific plant license renewal application.

Relating ASME Flaw Evaluation to ASME Task Group on Carbon Fiber Repairs

2017 ◽  
Author(s):  
Tomas Jimenez ◽  
Eric Houston ◽  
Nico Meyer
Keyword(s):  
Carbon Fiber ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Fiber Reinforced Polymers ◽  
Structural Factors ◽  
Operating Life ◽  
The Us ◽  
Piping Systems ◽  
Wave Passage ◽  
Regulatory Commission

As most nuclear power stations in the US have surpassed their initial 40 years of operability, the industry is now challenged with maintaining safe operations and extending the operating life of structures, systems and components. The US Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI), and Electric Power Research Institute (EPRI) have identified safety related buried piping systems as particularly susceptible to degradation. These systems are required to maintain the structural factors of the ASME Construction Codes under pressure and piping loads, which includes seismic wave passage. This paper focuses on evaluation approaches for metallic buried piping that can be used to demonstrate that localized thinning meets the requirements of the Construction Code. The paper then addresses a non-metallic repair option using carbon fiber reinforced polymers (CFRP) as the new pressure boundary.

Operational Events Noted for Power Uprated Plants and Potential Safety Implications

2002 ◽  
Author(s):  
A. W. Cronenberg ◽  
D. A. Powers ◽  
R. P. Savio
Keyword(s):  
Nuclear Regulatory Commission ◽  
System Component ◽  
Regulatory Requirements ◽  
Received Power ◽  
Potential Safety ◽  
Higher Power ◽  
The Us ◽  
Regulatory Commission ◽  
License Renewal ◽  
Control Rods

During the past several decades the US Nuclear Regulatory Commission (NRC) has reviewed and approved in excess of 50 licensee requests for power uprates, most of which have been in the range of 1–6% increase. More recently the agency has received License Amendment Requests (LARs) for significant power increases, in the range of 15–20%, which are under current review. Although each uprate request is evaluated to assure that current regulatory requirements are satisfied, concerns have developed regarding the safety implications of power uprates of this magnitude for the aging population of US-LWR plants. Such concerns stem from operational events noted for plants having received power uprates, including failure to fully-insert control rods in high-power/high-burnup PWR fuel assemblies, piping failures, and reactivity anomalies. Of particular concern is the potential for synergistic/compounding effects, for example higher core power when combined with system/component degradation via plant aging for plant license renewal, or higher power in conjunction with fuel life extensions to elevated burnup. This paper examines a number of operational events for uprated plants and potential safety implications of such events.

Method for Comparison of Different Material Types in Environmental Fatigue Screening

2020 ◽  
Author(s):  
Mark A. Gray ◽  
Jamie L. Oakman ◽  
Adam P. Walker
Keyword(s):  
Systematic Approach ◽  
Water Environment ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Metal Fatigue ◽  
Additional Consideration ◽  
Screening Process ◽  
Extended Operation ◽  
Regulatory Commission ◽  
License Renewal

Abstract Requirements for plant license renewal require the evaluation and management of time limited aging analyses for the period of extended operation. These include metal fatigue analyses, with additional consideration for the effects of the reactor water environment. Current license renewal requirements for evaluation of time limited aging analyses for fatigue have extended environmentally assisted fatigue (EAF) considerations to additional locations beyond the originally prescribed NUREG/CR-6260 locations. These requirements have resulted in implementation of a screening process to identify locations that are potentially more limiting for further evaluation of EAF. The screening process includes comparisons of fatigue locations within common bases, including consideration of applicable system transients, rigor of analysis methods, and material types. To date, the United States Nuclear Regulatory Commission has only accepted EAF screening evaluation results based on comparisons within similar material types. However, there are conditions under which a comparison of locations with different material types is valid. This paper presents a basis to compare locations having different material types, to potentially reduce the number of leading (“sentinel”) locations in EAF screening evaluations. The methodology is described through a systematic approach and basis to determine when locations are more limiting with respect to EAF when different material types are compared.

scholarly journals Review of Practice for Deeply Embedded/Buried NPP Structures Subject to Seismic Loadings

2004 ◽  
Author(s):  
J. Xu ◽  
C. Miller ◽  
C. Hofmayer ◽  
H. Graves
Keyword(s):  
National Laboratory ◽  
Earth Pressure ◽  
Nuclear Regulatory Commission ◽  
Brookhaven National Laboratory ◽  
Nuclear Containment ◽  
The Us ◽  
Reactor Building ◽  
Seismic Loadings ◽  
Knowledge And Practice ◽  
Regulatory Commission

Motivated by many design considerations, several conceptual designs for advanced reactors have proposed that the entire reactor building and a significant portion of the steam generator building will be either partially or completely embedded below grade. For the analysis of seismic events, the soil-structure interaction (SSI) effect and passive earth pressure for these types of deeply embedded structures will have a significant influence on the predicted seismic response. Sponsored by the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) is carrying out a research program to assess the significance of these proposed design features for advanced reactors, and to evaluate the existing analytical methods to determine their applicability and adequacy in capturing the seismic behavior of the proposed designs. This paper summarizes a literature review performed by BNL to determine the state of knowledge and practice for seismic analyses of deeply embedded and/or buried (DEB) nuclear containment type structures. Included in the paper is BNL’s review of the open literature of existing standards, tests, and practices that have been used in the design and analysis of DEB structures. The paper also provides BNL’s evaluation of available codes and guidelines with respect to seismic design practice of DEB structures. Based on BNL’s review, a discussion is provided to highlight the applicability of the existing technologies for seismic analyses of DEB structures and to identify gaps that may exist in knowledge and potential issues that may require better understanding and further research.

Severe Transportation Accidents: Impacts of Severe Fires on Radioactive Material Transportation

2009 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton ◽  
Darrell S. Dunn
Keyword(s):  
Spent Nuclear Fuel ◽  
Nuclear Regulatory Commission ◽  
Radioactive Material ◽  
Fire Investigation ◽  
Radioactive Materials ◽  
The Us ◽  
Material Transportation ◽  
Safe Transport ◽  
Regulatory Commission ◽  
Loss Of Strength

In 2007, a severe transportation accident occurred in Oakland, California in what is commonly known as the “MacArthur Maze” section of Interstate 580 (I-580). The accident involved a tractor trailer carrying gasoline that impacted an overpass support column and burst into flames. The subsequent fire burned for over 2 hours and led to the collapse of the overpass due to the loss of strength in the structural steel that supported the overpass. The US Nuclear Regulatory Commission (NRC) studied this accident to examine any potential regulatory implications related to the safe transport of radioactive materials, including spent nuclear fuel. This paper will discuss the details of the NRC’s MacArthur Maze fire investigation.

scholarly journals Assessment of CTF Boiling Transition and Critical Heat Flux Modeling Capabilities Using the OECD/NRC BFBT and PSBT Benchmark Databases

2013 ◽  
Vol 2013 ◽  
pp. 1-12
Author(s):  
Maria Avramova ◽  
Diana Cuervo
Keyword(s):  
Nuclear Energy ◽  
Nuclear Regulatory Commission ◽  
Nucleate Boiling ◽  
State University ◽  
Energy Agency ◽  
Fine Mesh ◽  
The Us ◽  
Penn State ◽  
Regulatory Commission ◽  
Rod Array

Over the last few years, the Pennsylvania State University (PSU) under the sponsorship of the US Nuclear Regulatory Commission (NRC) has prepared, organized, conducted, and summarized two international benchmarks based on the NUPEC data—the OECD/NRC Full-Size Fine-Mesh Bundle Test (BFBT) Benchmark and the OECD/NRC PWR Sub-Channel and Bundle Test (PSBT) Benchmark. The benchmarks’ activities have been conducted in cooperation with the Nuclear Energy Agency/Organization for Economic Co-operation and Development (NEA/OECD) and the Japan Nuclear Energy Safety (JNES) Organization. This paper presents an application of the joint Penn State University/Technical University of Madrid (UPM) version of the well-known sub-channel code COBRA-TF (Coolant Boiling in Rod Array-Two Fluid), namely, CTF, to the steady state critical power and departure from nucleate boiling (DNB) exercises of the OECD/NRC BFBT and PSBT benchmarks. The goal is two-fold: firstly, to assess these models and to examine their strengths and weaknesses; and secondly, to identify the areas for improvement.

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