Modifications of the 2016 Edition of the RCC-M Code to Account for Environmentally Assisted Fatigue

2016 ◽  
Author(s):  
Stéphan Courtin ◽  
Thomas Métais ◽  
Manuela Triay ◽  
Eric Meister ◽  
Stéphane Marie
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Steering Committee ◽  
Nuclear Industry ◽  
Life Extension ◽  
Asme Code ◽  
Nickel Base ◽  
Fatigue Evaluation ◽  
International Experts

The French nuclear industry has to face nowadays a series of challenges it did not have to face a decade ago. The most significant one is to ensure a reliable and safe operation of Nuclear Power Plants (NPP) in a context of both an ageing reactor fleet and new builds. The new constructions need rules that integrate a strong operation feedback while the older NPPs need rules that will guarantee the life extension beyond 40 years of operation. In this context, a new edition of the French RCC-M Code is planned for 2016. This new edition integrates the modifications made to the Code as a result of Requests for Modification (RM), which can be submitted by anyone and which help to continuously improve the quality and robustness of the Code. Concerning fatigue analyses, the RCC-M Code steering committee has acknowledged end of 2014 the reception of two RM to modify the fatigue design curve for austenitic stainless steels and Nickel base alloys, as well as to integrate environmental effects in the fatigue evaluation for austenitic stainless steel components. The contents of these two RM were based on the proposals presented in Reference [1]. AFCEN required a technical review of these two RM and this task was performed by a working group composed by French and international experts. This process concluded to the approval of these two RM to be integrated to the 2016 edition of the RCC-M Code. This paper offers a presentation of these two new Rules in Probation Phase (RPP), this format being quite similar to Code Cases proposed by ASME Code.

Fatigue Benchmark Comparison Effort Between Code_Aster and CNNC/NPIC Software: Part 2

2020 ◽  
Author(s):  
Hai Xie ◽  
Zichen Kong ◽  
Xuejiao Shao ◽  
Tanguy Mathieu ◽  
Furui Xiong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Austenitic Stainless Steels ◽  
Combination Method ◽  
Practical Implementation ◽  
Phase 2 ◽  
Software Comparison ◽  
Fatigue Design ◽  
Nickel Base ◽  
Fatigue Evaluation

Abstract Fatigue is identified as a significant degradation mode that affects nuclear power plants world-wide. Recent research on the interaction between fatigue degradation and the influence of PWR environment has caused international concern and triggered numerous research programs [1]. In this context, several codes & standards, including the RCC-M code, have included some technical mandatory or non-mandatory sections to address the issue. In RCC-M, this is compiled in the Rules in Probation Phase 2 and 3 [2]. Due to the novelty of these rules, there is room for improvement for the specific and practical implementation of these rules. AFCEN has hence launched a benchmark exercise at the end of 2019 to help increase the quality of these rules. Part 1 of this paper [3] states that EDF and CNNC/NPIC have launched an effort to benchmark their respective codes on fatigue calculation including the EAF algorithm. In the second part of the benchmark, the two companies started the code comparison based on a benchmark case provided by AFCEN. As stated previously, the 2016 edition of RCC-M code integrates the modifications made to the Code in Probation Phase 2 and 3(RPP)[2], which respectively modify the fatigue design curve for austenitic stainless steels and Nickel base alloys, as well as integrate environmental effects in the fatigue evaluation for austenitic stainless steel components. In this paper, a comparison between RCC-M RPP and NUREG/CR-6909 rev.1 [3] is proposed. The comparison focuses on the technical details of the strain rate calculation and transient combination method. The cumulative fatigue usage factor with or without considering EAF according to RCC-M RPP – 2 and RPP – 3 is given by EDF, using code_aster and its POST_RCCM operator. CNNC/NPIC will provide multiple sets of results including cumulative fatigue usage factors according to RCC-M RPP and NUREG CR/6909 rev. 1 respectively using its own software. Comparison of selection for peak and valleys points, Sn and Fen values are also presented. Differences of the algorithms of the two codes are also discussed.

Fatigue Monitoring System and Post-Processing of Temperature Measurements: Surge Line Under Stratification Loading

2015 ◽  
Author(s):  
Benoit Jouan ◽  
Jürgen Rudolph ◽  
Steffen Bergholz
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Thermal Loading ◽  
Austenitic Stainless Steels ◽  
Nuclear Industry ◽  
Integral Approach ◽  
Safety Standards ◽  
Analysis Process ◽  
Surge Line ◽  
Fatigue Evaluation

The ageing management of power plants is nowadays a main issue for all nuclear industry actors: states, regulatory agencies, operators, designers or suppliers. Consequently, lots of operators have to deal with demanding safety requirements to ensure the operation of power plants particularly in the context of lifetime extension. With regard of the fatigue assessment of nuclear components, stringent safety standards are synonymous of new parameters to take into account in the fatigue analysis process such as for instance: new design of fatigue curves particularly for austenitic stainless steels, the consideration of environmentally assisted fatigue (EAF) and stratification effects. In this context AREVA developed within the integral approach AREVA Fatigue Concept (AFC) new tools and methods to live up to operators expectations. The last mentioned stratification issue will be focused on in the framework of this dedicated paper. Based on measured thermal loads, the Fast Fatigue Evaluation (FFE) process allows for highly-automated and reliable data processing to evaluate time-dependent cumulative usage factors of mechanical components. This method has recently been extended to the consideration of stratification loading with surge line application. The paper presents the latest AREVA research and development activities on the FFE method applied to a surge line under stratification thermal loading. An additional CFD analysis was performed in order to calculate realistic thermal loadings during start-up conditions of nuclear power plant conditions. The FFE methodology was used to calculate thermal stress at all relevant locations. This approach opens the possibility of a realistic CUF calculation. The methodology, the principle results and benefits are presented in the paper.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping-equipment-structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.

Comparison of ASME Code NB-3200 and NB-3600-Based Environmental Fatigue Calculations for Piping Components

2014 ◽  
Author(s):  
Kyeongjin Yang ◽  
Daesu Kim ◽  
Dongjae Lee ◽  
Joonho Lee ◽  
Sangbae Lee ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Time Consistency ◽  
Correction Factors ◽  
Pressurized Water ◽  
Water Reactor ◽  
Practical Applications ◽  
Thermal Transients ◽  
Asme Code ◽  
Fatigue Evaluation

Environmentally-assisted fatigue evaluations are to be conducted for ASME Code Class 1 piping components in a pressurized water reactor. Environmental fatigue correction factor method for incorporating the effects of light water reactor coolant environments into ASME Section III fatigue evaluations was investigated in this paper. Both ASME Code NB-3200 and NB-3600 methods were used to determine the usage factors of the piping components. Considered in these calculations were the loads which are generally applied to the piping design for the nuclear power plants such as seismic, thermal expansions, thermal transients, thermal stratifications and building-filtered dynamic loadings. For the practical applications of NB-3600 method, regarded as the simple and conservative approach, to the piping components, it was presumed that the stress intensity and/or strain time histories for all or some of the external loadings were not known; therefore the time consistency might not be considered in calculating the usage factors as well as environmental correction factors (Fen). In NB-3200 method in contrast to NB-3600, the stress variations with time for all loads except for the dynamic loads were obtained for the fatigue evaluations in LWR environments, and therefore the time consistency was considered. The results showed that the environmental correction factors as well as in-air cumulative usage factors calculated from NB-3200 methods were significantly less than those from NB-3600 rules. In addition, comparing the results of conventional ASME fatigue evaluation applied until 2006 to the ones in accordance with USNRC RG 1.207 issued on 2007, one may identify that the cumulative usage factors in LWR environments were larger than the conventional one due to the change of design fatigue curves as well as Fen factors accounting for the environmental effects on fatigue. Although this work was focused on the detailed calculations of the usage factors and Fen values, one might identify or suggest a number of areas requiring further clarification or research through the efforts of this study, which were not yet addressed. A few items needed to be clarified, especially for NB-3600-based fatigue evaluations, are also discussed in this paper.

An Investigation on the Reduction Schemes of Environmental Fatigue Usage Factor for Pressurizer Surge Line

2011 ◽  
Author(s):  
Sun-yeh Kang ◽  
Won-ho Jo ◽  
Min-sup Song ◽  
Ki-seok Yoon ◽  
Taek-sang Choi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Light Water Reactor ◽  
Life Extension ◽  
Regulatory Requirement ◽  
Light Water ◽  
Plant Life ◽  
Asme Code ◽  
Surge Line ◽  
Critical Components ◽  
Fatigue Evaluation

For plant life extension, it is the regulatory requirement to assess reactor coolant environmental impacts on critical components of the nuclear power plant including at least those mentioned in NUREG/CR-6260[2]. The pressurizer surge line is the most easy-to-fail component in view of LWR (Light Water Reactor) environments when it comes to meeting the current ASME code limit of the fatigue evaluation. Cumulative Usage Factor (CUF) value could be increased to a maximum of 15.35 times due to the environmental effects, which makes it easy to exceed the allowable fatigue limit (1.0). This paper discusses the process of the environmental correction factor calculation described in NUREG/CR-5704[4], and five proposed schemes for reducing the environmental CUF value to the ASME code limit or below. This paper concludes that the proposed schemes are effective in lowering the environmental CUF value of the pressurizer surge line.

A Review of the Effects of Coolant Environments on the Fatigue Life of LWR Structural Materials

2009 ◽  
Vol 131 (2) ◽  
Author(s):  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Fatigue Life ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Crack Initiation ◽  
Austenitic Stainless Steels ◽  
Structural Materials ◽  
Low Alloy Steels ◽  
Asme Code ◽  
Alloy Steels

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code specifies design curves for the fatigue life of structural materials in nuclear power plants. However, the effects of light water reactor (LWR) coolant environments were not explicitly considered in the development of the design curves. The existing fatigue-strain-versus-life (ε-N) data indicate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. Under certain environmental and loading conditions, fatigue lives in water relative to those in air can be a factor of 15 lower for austenitic stainless steels and a factor of ≈30 lower for carbon and low-alloy steels. This paper reviews the current technical basis for the understanding of the fatigue of piping and pressure vessel steels in LWR environments. The existing fatigue ε-N data have been evaluated to identify the various material, environmental, and loading parameters that influence fatigue crack initiation and to establish the effects of key parameters on the fatigue life of these steels. Statistical models are presented for estimating fatigue life as a function of material, loading, and environmental conditions. An environmental fatigue correction factor for incorporating the effects of LWR environments into ASME Code fatigue evaluations is described. This paper also presents a critical review of the ASME Code fatigue design margins of 2 on stress (or strain) and 20 on life and assesses the possible conservatism in the current choice of design margins.

Regulatory Requirements and Perspectives on the Use of Inservice Testing Software in the Nuclear Industry

2014 ◽  
Author(s):  
Jason B. Carneal
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Industry ◽  
Light Water Reactor ◽  
Federal Regulations ◽  
Regulatory Requirements ◽  
Operational Readiness ◽  
Asme Code ◽  
American Society

The American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code) establishes the requirements for preservice and inservice testing (IST) and examination of certain components to assess their operational readiness in light-water reactor nuclear power plants. The Code of Federal Regulations (CFR) endorses the use of the ASME OM Code in 10 CFR 50.55a(b)(3) . This paper focuses on applicable regulatory requirements and regulatory perspectives associated with the use of IST software in the nuclear industry. Paper published with permission.

Life Extension of Nuclear Power Plants: World Situation and the USA Case

2010 ◽  
pp. 50-56 ◽  
Author(s):  
Pablo T. León ◽  
Loreto Cuesta ◽  
Eduardo Serra ◽  
Luis Yagüe
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Life Extension ◽  
The Usa

scholarly journals An Approach to Analyze Diagnosis Errors in Advanced Main Control Room Operations Using the Cause-Based Decision Tree Method

Energies ◽  
2021 ◽  
Vol 14 (13) ◽  
pp. 3832
Author(s):  
Awwal Mohammed Arigi ◽  
Gayoung Park ◽  
Jonghyun Kim
Keyword(s):  
Decision Tree ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Cognitive Activity ◽  
Emergency Operation ◽  
Nuclear Industry ◽  
Working Environment ◽  
Control Rooms ◽  
Diagnosis Errors

Advancements in the nuclear industry have led to the development of fully digitized main control rooms (MCRs)—often termed advanced MCRs—for newly built nuclear power plants (NPPs). Diagnosis is a major part of the cognitive activity in NPP MCRs. Advanced MCRs are expected to improve the working environment and reduce human error, especially during the diagnosis of unexpected scenarios. However, with the introduction of new types of tasks and errors by digital MCRs, a new method to analyze the diagnosis errors in these new types of MCRs is required. Task analysis for operator diagnosis in an advanced MCR based on emergency operation was performed to determine the error modes. The cause-based decision tree (CBDT) method—originally developed for analog control rooms—was then revised to a modified CBDT (MCBDT) based on the error mode categorizations. This work examines the possible adoption of the MCBDT method for the evaluation of diagnosis errors in advanced MCRs. We have also provided examples of the application of the proposed method to some common human failure events in emergency operations. The results show that with some modifications of the CBDT method, the human reliability in advanced MCRs can be reasonably estimated.

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.

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