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.

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.

Automatic Fatigue Monitoring Based on Real Loads - Calculation Example of a Flange

2013 ◽  
Author(s):  
Benoît Jouan ◽  
Steffen Bergholz ◽  
Jürgen Rudolph ◽  
Günter König ◽  
Andreas Manke
Keyword(s):  
Power Plants ◽  
Environmental Parameters ◽  
Nuclear Industry ◽  
Security Requirements ◽  
Integral Approach ◽  
Safety Standards ◽  
Calculation Methodology ◽  
Fatigue Evaluation ◽  
Ageing Management ◽  
Operating Cycles

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 security requirements to ensure the operation of power plants. Regarding with fatigue assessment of nuclear components, stringent safety standards are synonymous of new parameters to take into account in the fatigue analysis process, for instance: new design of fatigue curves, consideration of environmental parameters or 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. Based on measured thermal loads, the Fast Fatigue Evaluation (FFE) process allows for highly-automated and reliable data processing to evaluate time-dependant cumulative usage factors of mechanical components. Calculation and management of results are performed with the software FAMOSi, thus impact of operating cycles on components in terms of stress but also with regard of fatigue can be taken into account to plan an optimized decision related to the plant operation or maintenance activities. The FFE process was exemplary applied in 2012 in the EnBW Power Plant of Neckarwestheim (GKN II) to perform an informative fatigue diagnose of a spray line flange for different operating cycles. This paper describes the calculation methodology but also some relevant results to point out the benefits of this method to the ageing management of mechanical parts.

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.

Towards European Regulations for Nuclear Installations: Contribution of Nuclear Licensees

2009 ◽  
Author(s):  
Werner Zaiss
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Legal Framework ◽  
Nuclear Industry ◽  
Decision Makers ◽  
Safety Standards ◽  
Harmonization Process ◽  
European Regulations ◽  
Expert Input ◽  
European Nuclear

The European nuclear industry recognises that the liberalisation of the European energy market has led to the deregulation of electricity generation and supply and that diversity of national regulations could seriously distort competition. Undoubtedly, harmonizing regulations is the best way of ensuring that the industry can evolve within a stable legal framework. Consequently, nuclear license holders supported the work of the Western European Nuclear Regulators Association (WENRA) on the harmonization of European safety standards for existing nuclear power plants, as well as for waste and decommissioning. This support led to the creation, within FORATOM, of the ENISS (European Nuclear Installations Safety Standards) Initiative, in May 2005, in Brussels. The principal mission of ENISS is to bring together decision-makers, operators and specialists from the nuclear industry with national regulators in order to identify and possibly agree upon the scope and substance of harmonized safety standards. ENISS currently represents the nuclear utilities and operating companies from 17 European countries with nuclear power programme. ENISS above all provides the nuclear industry with the platform that it needs to express its views, provide expert input and interact fully with regulators throughout the harmonization process. ENISS first task has been to present a common industry position with regards to the Safety Reference Levels that WENRA has proposed. By engaging in constructive debate with WENRA and playing a dynamic role in the process, ENISS also defends the industry’s interests in a proactive way. The work of ENISS is a good example of how dialogue and results-oriented participation with stakeholders can help identify optimal solutions to the problems that our industry faces today. Another task of ENISS is to strengthen the industry influence in the revision work of the IAEA Safety Standards as well as in the European Directive on Nuclear Safety.

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.

Architecting “Maintenance Program” for Piping and Pipe Support

2013 ◽  
Author(s):  
Koichi Tsumori ◽  
Yoshizumi Fukuhara ◽  
Hiroyuki Terunuma ◽  
Koji Yamamoto ◽  
Satoshi Momiyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Maintenance Management ◽  
3D Cad ◽  
Cad System ◽  
Nuclear Disaster ◽  
Safety Regulations ◽  
Maintenance Program

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations. The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.

Research Progress on Radiation Shielding Concrete

2012 ◽  
Vol 253-255 ◽  
pp. 303-307 ◽  
Author(s):  
Jing Yang ◽  
Zhen Fu Chen ◽  
Yuan Chu Gan ◽  
Qiu Wang Tao
Keyword(s):  
Nuclear Waste ◽  
Material Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Radiation Shielding ◽  
Nuclear Industry ◽  
Research Progress ◽  
Concrete Durability ◽  
Technology Research ◽  
Shielding Material

Radiation shielding concrete is widely used in nuclear power plants, accelerators, hospitals, etc. With the development of nuclear industry technology, research on radiation shielding material properties is of great importance. Research on properties of radiation shielding concrete with different aggregates or admixtures and the effect of high temperature on the performance of shielding concrete are introduced. Along with the nuclear waste increase, shielding concrete durability and nuclear waste disposal are getting paramount.

Slow Crack Growth Resistance of Parent and Joint Materials From PE4710 Piping for Safety-Related Nuclear Power Plant Piping

2011 ◽  
Author(s):  
S. Kalyanam ◽  
D.-J. Shim ◽  
P. Krishnaswamy ◽  
Y. Hioe
Keyword(s):  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Slow Crack Growth ◽  
Nuclear Industry ◽  
Pipe Material ◽  
Service Water ◽  
Hdpe Pipe ◽  
Pipe Materials ◽  
Hdpe Pipes

HDPE pipes are considered by the nuclear industry as a potential replacement option to currently employed metallic piping for service-water applications. The pipes operate under high temperatures and pressures. Hence HDPE pipes are being evaluated from perspective of design, operation, and service life requirements before routine installation in nuclear power plants. Various articles of the ASME Code Case N-755 consider the different aspects related to material performance, design, fabrication, and examination of HDPE materials. Amongst them, the material resistance (part of Article 2000) to the slow crack growth (SCG) from flaws/cracks present in HDPE pipe materials is an important concern. Experimental investigations have revealed that there is a marked difference (almost three orders less) in the time to failure when the notch/flaw is in the butt-fusion joint, as opposed to when the notch/flaw is located in the parent HDPE material. As part of ongoing studies, the material resistance to SCG was investigated earlier for unimodal materials. The current study investigated the SCG in parent and butt-fusion joint materials of bimodal HDPE (PE4710) pipe materials acquired from two different manufacturers. The various stages of the specimen deformation and failure during the creep test are characterized. Detailed photographs of the specimen side-surface were used to monitor the specimen damage accumulation and SCG. The SCG was tested using a large specimen (large creep frame) as well as using a smaller size specimen (PENT frame) and the results were compared. Further, the effect of polymer orientation or microstructure in the bimodal HDPE pipe on the SCG was studied using specimens with axial and circumferential notch orientations in the parent pipe material.

scholarly journals Nuclear Power: Time to Start Again

2003 ◽  
Author(s):  
William D. Rezak
Keyword(s):  
Power Generation ◽  
Electric Power ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Environmentally Friendly ◽  
Power Industry ◽  
Nuclear Industry ◽  
The Us ◽  
Generating Capacity

One of America’s best kept secrets is the success of its nuclear electric power industry. This paper presents data which support the construction and operating successes enjoyed by energy companies that operate nuclear power plants in the US. The result—the US nuclear industry is alive and well. Perhaps it’s time to start anew the building of nuclear power plants. Let’s take the wraps off the major successes achieved in the nuclear power industry. Over 20% of the electricity generated in the United States comes from nuclear power plants. An adequate, reliable supply of reasonably priced electric energy is not a consequence of an expanding economy and gross national product; it is an absolute necessity before such expansion can occur. It is hard to imagine any aspect of our business or personal lives not, in some way, dependent upon electricity. All over the world (in 34 countries) nuclear power is a low-cost, secure, safe, dependable, and environmentally friendly form of electric power generation. Nuclear plants in these countries are built in six to eight years using technology developed in the US, with good performance and safety records. This treatise addresses the success experienced by the US nuclear industry over the last 40 years, and makes the case that this reliable, cost-competitive source of electric power can help support the economic engine of the country and help prevent experiences like the recent crisis in California. Traditionally, the evaluation of electric power generation facility performance has focused on the ability of plants to produce at design capacity for high percentages of the time. Successful operation of nuclear facilities is determined by examining capacity or load factors. Load factor is the percentage of design generating capacity that a power plant actually produces over the course of a year’s operation. This paper makes the case that these operating performance indicators warrant renewed consideration of the nuclear option. Usage of electricity in the US now approaches total generating capacity. The Nuclear Regulatory Commission has pre-approved construction and operating licenses for several nuclear plant designs. State public service commissions are beginning to understand that dramatic reform is required. The economy is recovering and inflation is minimal. It’s time, once more, to turn to the safe, reliable, environmentally friendly nuclear power alternative.

Sign in / Sign up

Export Citation Format

Share Document