Research on the Application of Fen for Environmentally Assisted Fatigue Evaluation of the Austenitic SS Pipe Under Combined Transient Loads

2018 ◽  
Vol 4 (4) ◽  
Author(s):  
Bingbing Liang ◽  
Xu Zhang ◽  
Haifeng Yin ◽  
Yang Dai
Keyword(s):  
Fatigue Life ◽  
Research Work ◽  
Nuclear Regulatory Commission ◽  
Combined Loading ◽  
Reactor Materials ◽  
Fatigue Evaluation ◽  
Regulatory Commission ◽  
Conservative Result ◽  
Fatigue Life Reduction ◽  
Temperature Strain

Accumulative test data indicate that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of primary pressure boundary components materials to be significantly reduced. Environmentally assisted fatigue (EAF) is the abbreviation of the environmentally assisted fatigue. In 2007, Nuclear Regulatory Commission (NRC) issued RG. 1.207. It was updated in 2014. And, it requires that the effects of LWR environment on the fatigue life reduction of metal components should be considered for new design plants. And it suggests to use environmental correction factor, Fen, to account for EAF. NRC regulation (NUREG), NUREG/CR-6909 (NRC, 2013, “Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials,” U.S. Nuclear Regulatory Commission, Argonne, IL, Standard no. NUREG/CR-6909), presents the detail Fen calculation formula. Fen is a function of temperature, strain rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depending on the experience of the primary pressure boundary piping transient operation, Fen varies during each transient. More uncertainty and confusion are raised during the application of the Fen method. The research work in this paper includes: first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depending on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multitransient loading conditions. And the results are compared with those of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test plan is presented.

Research on the Application of FEN for EAF Evaluation of the Austenitic SS Pipe Under Combined Transient Loads

2017 ◽  
Author(s):  
Bingbing Liang ◽  
Xu Zhang ◽  
Haifeng Yin ◽  
Yang Dai
Keyword(s):  
Research Work ◽  
Water Environment ◽  
Combined Loading ◽  
Light Water Reactor ◽  
Light Water ◽  
Thermal Transient ◽  
Materials Used ◽  
Conservative Result ◽  
Fatigue Life Reduction ◽  
Temperature Strain

Accumulative test data indicates that the effects of the light water reactor (LWR) environment could cause the fatigue resistance of materials used in the reactor coolant pressure boundary components to significantly reduce. EAF is used as the abbreviation of the environmentally assisted fatigue in the nuclear field. In 2007, NRC issued RG. 1.207. It was updated in 2014. And it requires that the effects of the light-water environment on the fatigue life reduction of metal components should be considered for new plants. And it suggests to use environmental correction factor (Fen) to account for EAF. Fen = Nair/Nwater (N is occurrences). NUREG/CR-6909 [1] presents the detail Fen calculation formula which includes the complicated influence of combined multi-parameters. Fen is a function of temperature, strain amplitude & rate, dissolved oxygen level in water, and sulfur content of the steel. Accordingly, Fen calculation will present a comparatively conservative result. Depends on the experience of the primary pressure boundary piping transient operation, Fen vary during each transient. More uncertainty and confusion are raised during the application of the Fen method. In the research work involved in this article, first, the typical character of piping thermal transient is derived based on the existing experience. Second, small specimen EAF tests are conducted depend on the above derived combined loading characters. Then effort is taken to improve the application of the Fen method for the combined multi-transient loading conditions. And the result is compared with that of the lowest instantaneous Fen method and equalization of the weighted Fen method. Finally, a designed test matrix is needed to prove its practicability furthermore.

Comparison of Component Fatigue Test Data With Proposed EAF Methodologies for Austenitic Stainless Steel

2009 ◽  
Author(s):  
Rim Nayal ◽  
Hasan Charkas
Keyword(s):  
Stainless Steel ◽  
Fatigue Life ◽  
Fatigue Test ◽  
Low Cycle Fatigue ◽  
Nuclear Regulatory Commission ◽  
Fatigue Curve ◽  
Asme Code ◽  
Regulatory Commission ◽  
License Renewal ◽  
The Impact

The U.S. Nuclear Regulatory Commission (NRC) currently requires evaluation of the effect of environmental fatigue for both license renewal and new plants. NRC required the use of methodology in EPRI MRP-47, Rev. 1 addressing NUREG/CR-5704, be used for license renewal of stainless steel (SS) components, and NUREG/CR-6909 for use in new plants. These two methodologies are based on applying an environmental correction factor (Fen) on the number of in-air design cycles. These factors are applied to the fatigue usage from each individual range of stress (or range of strain). The focus of this paper is to compare the two aforementioned methodologies; this includes comparison of the fatigue curve as well as the comparison of the environmental correction factors (Fen). Fatigue test results data reported by others are also compared with these two methodologies. It is important to evaluate the impact of using any of those methodologies on the design fatigue life of the components. It is concluded that NUREG/CR-5704 is more severe than NUREG/CR-6909 in the LCF (low-cycle fatigue) regime, while NUREG/CR-6909 is more severe elsewhere, and both NUREG’s extremely underestimate fatigue life in PWR environment. It is also concluded that the current ASME-code fatigue curve for stainless steel reasonably estimates fatigue life in an LWR environment with reasonable margins.

Stress-Based Environmental Fatigue Monitoring of PWR Charging Nozzle

2013 ◽  
Author(s):  
Timothy Gilman ◽  
Archana Chinthapalli ◽  
Michael Hoehn
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water ◽  
Technical Report ◽  
Fatigue Monitoring ◽  
Energy Center ◽  
Plant Data ◽  
Water Reactors ◽  
Fatigue Evaluation ◽  
Regulatory Commission

This paper describes the techniques utilized to perform a stress-based environmentally-assisted fatigue evaluation of Westinghouse-designed charging branch nozzles on the reactor coolant loop of the Callaway Energy Center nuclear power plant. Analysis results from using idealized, design transient definitions are compared to those resulting from analysis of the actual plant data. Benchmarking analyses, performed to address Nuclear Regulatory Commission (NRC) concerns about simplified methodologies, are described. The simplified results are also compared to those produced using an advanced, multiaxial stress-based fatigue methodology defined in a recent EPRI technical report [3]. This paper concludes that stress-based fatigue monitoring using actual plant data is an effective way for a plant to manage environmentally-assisted fatigue of charging nozzles in pressurized water reactors (PWRs).

Applicability of the Modified Rate Approach Method Under Various Conditions Simulating Actual Plant Conditions

2006 ◽  
Author(s):  
Katsumi Sakaguchi ◽  
Yuichiro Nomura ◽  
Shigeki Suzuki ◽  
Hiroshi Kanasaki
Keyword(s):  
Fatigue Life ◽  
Strain Rate ◽  
Fatigue Tests ◽  
Transient Condition ◽  
Stress Cycle ◽  
Rate Condition ◽  
Approach Method ◽  
Strain Rate Condition ◽  
Fatigue Evaluation ◽  
Fatigue Life Reduction

The fatigue life in elevated temperature water is strongly affected by water chemistry, temperature and strain rate. The effects of these parameters on fatigue life reduction have been investigated experimentally. In transient condition in an actual plant, however, such parameters as temperature and strain rate are not constant. In order to evaluate fatigue damage in actual plant on the basis of experimental results under constant temperature and strain rate condition, the modified rate approach method was developed. As a part of the EFT (Environmental Fatigue Tests) project, the study was conducted in order to evaluate the applicability of the modified rate approach to the case where temperature and strain rate varied simultaneously. It was reported in the previous papers (1,2) that the accuracy of modified rate approach is about factor of 2. Various kinds of transient have to be taken into account of in actual plant fatigue evaluation, and stress cycle of several ranges of amplitude has to be considered in assessing damage from fatigue. Generally, cumulative usage factor is applied in this type of evaluation. In this study, in order to confirm applicability of modified rate approach method together with cumulative usage factor, tests were carried out by combining stress cycle blocks of different strain amplitude levels, in which temperature changes in response to strain change in a simulated PWR environment.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

Use of HFE Guidance for the Review of Nuclear Power Plants

2020 ◽  
Vol 64 (1) ◽  
pp. 1-5
Author(s):  
John O’Hara ◽  
Stephen Fleger
Keyword(s):  
Interface Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Nuclear Regulatory Commission ◽  
Program Review ◽  
Human Factors Engineering ◽  
Plant Design ◽  
Review Model ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) evaluates the human factors engineering (HFE) of nuclear power plant design and operations to protect public health and safety. The HFE safety reviews encompass both the design process and its products. The NRC staff performs the reviews using the detailed guidance contained in two key documents: the HFE Program Review Model (NUREG-0711) and the Human-System Interface Design Review Guidelines (NUREG-0700). This paper will describe these two documents and the method used to develop them. As the NRC is committed to the periodic update and improvement of the guidance to ensure that they remain state-of-the-art design evaluation tools, we will discuss the topics being addressed in support of future updates as well.

The MacArthur Maze Fire and Roadway Collapse: A “Worst Case Scenario” for Spent Nuclear Fuel Transportation?

2012 ◽  
Author(s):  
Christopher S. Bajwa ◽  
Earl P. Easton ◽  
Harold Adkins ◽  
Judith Cuta ◽  
Nicholas Klymyshyn ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Steel Structure ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Severe Accident ◽  
Case Scenario ◽  
Worst Case ◽  
Work Related ◽  
Regulatory Commission

In 2007, a severe transportation accident occurred near Oakland, California, at the interchange known as the “MacArthur Maze.” The accident involved a double tanker truck of gasoline overturning and bursting into flames. The subsequent fire reduced the strength of the supporting steel structure of an overhead interstate roadway causing the collapse of portions of that overpass onto the lower roadway in less than 20 minutes. The US Nuclear Regulatory Commission has analyzed what might have happened had a spent nuclear fuel transportation package been involved in this accident, to determine if there are any potential regulatory implications of this accident to the safe transport of spent nuclear fuel in the United States. This paper provides a summary of this effort, presents preliminary results and conclusions, and discusses future work related to the NRC’s analysis of the consequences of this type of severe accident.

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.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

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