Aging Management Program of Large Diameter Safety Class Piping

2006 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Diameter ◽  
Thermal Ageing ◽  
Management Program ◽  
Lessons Learned ◽  
Energy Agency ◽  
Ageing Management ◽  
Fatigue Corrosion ◽  
Systematic Program

Ageing management of Nuclear Power Plants is an essential issue for utilities, in term of safety and availability and corresponding economical consequences. Practically all nuclear countries have developed a systematic program to deal with ageing of components on their plants. This paper presents the ageing management program developed by EDF (Electricite´ de France) that are compared with different approaches in other countries, like IAEA (International Atomic Energy Agency)guidelines and GALL (Generic Ageing Lessons Learned) report. The paper presents an example of application to large diameter safety class piping. Different degradation mechanisms are considered fatigue, corrosion and thermal ageing. Maintenance and surveillance actions are also discussed in the paper.

Aging Management Program of Large Diameter Safety Class Piping

2005 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Large Diameter ◽  
Management Program ◽  
Degradation Mechanisms ◽  
Ageing Management ◽  
Fatigue Corrosion ◽  
Systematic Program

Ageing management of Nuclear Power Plants is an essential issue for utilities, in term of safety and availability and corresponding economical consequences. Practically all nuclear countries have developed a systematic program to deal with ageing of components on their plants. This paper presents the ageing management program developed by EDF and that are compared with different other approaches in other countries (IAEA guidelines and GALL report). The paper presents an example of application to large diameter safety class piping. Different degradation mechanisms are considered like fatigue, corrosion or thermal aging. Maintenance and surveillance actions are discussed in the paper.

Research of Screening and Grading Method of Ageing Sensitive Objectives in Nuclear Power Plant

2014 ◽  
Vol 543-547 ◽  
pp. 858-861
Author(s):  
Xiao Tian Liu ◽  
Yong Wang ◽  
Shao Rui Niu ◽  
Yan Zhao Zhang ◽  
Zhen Hao Shi ◽  
...  
Keyword(s):  
Fuzzy Logic ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Lessons Learned ◽  
Nonlinear Coupling ◽  
Highly Nonlinear ◽  
Ageing Management ◽  
First Time

This first step of ageing management in nuclear power plant is to determine the objectives and their priorities. The characteristics of the objectives are complex and highly nonlinear coupling. A fuzzy logic based screening and grading method have been developed in this research for the first time which combined the genetic ageing lessons learned and field expert experience to resolve the problem. The method have been approved of highly applicability and applied to ageing management in multiple nuclear power plants.

scholarly journals Demonstration of Fatigue for LTO License of NPP Borssele

2015 ◽  
Author(s):  
M. H. C. Hannink ◽  
F. J. Blom ◽  
P. W. B. Quist ◽  
A. E. de Jong ◽  
W. Besuijen
Keyword(s):  
Environmental Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Integrated Approach ◽  
Temperature Measurements ◽  
Fatigue Assessment ◽  
Term Operation ◽  
Load Monitoring ◽  
Number Of Cycles ◽  
Ageing Management

Long Term Operation (LTO) of nuclear power plants (NPPs) requires an ageing management review and a revalidation of Time Limited Ageing Analyses (TLAAs) of structures and components important for nuclear safety. An important ageing effect to manage is fatigue. Generally, the basis for this is formed by the fatigue analyses of the safety relevant components. In this paper, the methodology for the revalidation of fatigue TLAAs is demonstrated for LTO of NPP Borssele in the Netherlands. The LTO demonstration starts with a scoping survey to determine the components and locations having relevant fatigue loadings. The scope was defined by assessment against international practice and guidelines and engineering judgment. Next, a methodical review was performed of all existing fatigue TLAAs. This also includes the latest international developments regarding environmental effects. In order to reduce conservatism, a comparison was made between the number of cycles in the analyses and the number of cycles projected to the end of the intended LTO period. The projected number of cycles is based on transient counting. The loading conditions used in the analyses were assessed by means of temperature measurements by the fatigue monitoring system (FAMOS). As a result of the review, further fatigue assessment or assessment of environmental effects was necessary for certain locations. New analyses were performed using state-of-the-art calculation and assessment methods. The methodology is demonstrated by means of an example of the surge line. The model includes the piping, as well as the nozzles on the pressurizer and the main coolant line. The thermal loadings for the fatigue analysis are based on temperature measurements. Fatigue management of the NPP is ensured by means of the fatigue concept where load monitoring, transient counting and fatigue assessment are coupled through an integrated approach during the entire period of LTO.

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.

Transition to Online Cable Insulation Condition Monitoring

2021 ◽  
Author(s):  
S. W. Glass ◽  
Leonard S. Fifield ◽  
Mychal P. Spencer
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Management Program ◽  
Capacitance Measurement ◽  
Test Bed ◽  
Test Environment ◽  
Specific Test ◽  
Motor Test

Abstract Nuclear power plant cables were originally qualified for 40 year life and generally have not required specific test verification to assure service availability through the initial plant qualification period. However, license renewals to 60 and 80 years of operation require a cable aging management program that depends on some form of test and verification to assure fitness for service. Environmental stress (temperature, radiation, chemicals, water, and mechanical) varies dramatically within a nuclear power plant and, in some cases, cables have degraded and required repair or replacement before their qualified end-of-life period. In other cases, cable conditions have been mild and dependable cable performance confirmed to extend well beyond the initial qualified life. Most offline performance-based testing requires cables to be decoupled and de-energized for specially trained technicians to perform testing. These offline tests constitute an expensive operational burden that limits the economic viability of nuclear power plants. Although initial investment may be higher, new online test practices are emerging as options or complements to offline testing that avoid or minimize the regularly scheduled offline test burden. These online methods include electrical and fiber-optic partial discharge measurement, spread spectrum time or frequency domain reflectometry, distributed temperature profile measurements, and local interdigital capacitance measurement of insulation characteristics. Introduction of these methods must be supported by research to confirm efficacy plus either publicly financed or market driven investment to support the start-up expense of cost-effective instrumentation to monitor cable condition and assure reliable operation. This work summarizes various online cable assessment technologies plus introduces a new cable motor test bed to assess some of these technologies in a controlled test environment.

scholarly journals Study on Probabilistic Safety Goals for Multimodule High-Temperature Gas-Cooled Reactor Based on Chinese Societal Risks

2021 ◽  
Vol 2021 ◽  
pp. 1-10
Author(s):  
Jinghan Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
Nuclear Safety ◽  
Energy Agency ◽  
Water Reactor ◽  
Regulatory Commission ◽  
High Temperature Gas ◽  
Probabilistic Safety

Nuclear safety goal is the basic standard for limiting the operational risks of nuclear power plants. The statistics of societal risks are the basis for nuclear safety goals. Core damage frequency (CDF) and large early release frequency (LERF) are typical probabilistic safety goals that are used in the regulation of water-cooled reactors currently. In fact, Chinese current probabilistic safety goals refer to the Nuclear Regulatory Commission (NRC) and the International Atomic Energy Agency (IAEA), and they are not based on Chinese societal risks. And the CDF and LERF proposed for water reactor are not suitable for high-temperature gas-cooled reactors (HTGR), because the design of HTGR is very different from that of water reactor. And current nuclear safety goals are established for single reactor rather than unit or site. Therefore, in this paper, the development of the safety goal of NRC was investigated firstly; then, the societal risks in China were investigated in order to establish the correlation between the probabilistic safety goal of multimodule HTGR and Chinese societal risks. In the end, some other matters about multireactor site were discussed in detail.

Information Technology and the Development of a Global Safety Culture

2011 ◽  
pp. 558-581
Author(s):  
Susan L. Rothwell
Keyword(s):  
Information Technology ◽  
Safety Culture ◽  
Nuclear Power ◽  
Power Plants ◽  
Health And Safety ◽  
Lessons Learned ◽  
Energy Independence ◽  
Power Safety ◽  
Under Construction ◽  
Culture Development

A nuclear power plant is one of the most complex sociotechnical systems ever created, with operation requiring multiple organizations, extensive interaction, and a mission to protect public health and safety. A strong global nuclear power safety culture is important, with over 400 nuclear power plants worldwide and more under construction to reduce fossil fuel dependency. We increasingly rely on technology, stressing our need for energy independence, security, reliability, education, and safety. Lessons learned from nuclear power safety culture development have a large potential audience. Unfortunately, the complexity of nuclear power and restricted access to operational data have limited outside research on and understanding of nuclear power safety culture. This chapter provides a conceptual, methodological, empirical, and operational perspective on the development of commercial nuclear power safety culture, focusing on the role of information technology (IT) in building, maintaining, and expanding global nuclear power safety culture.

scholarly journals Reactor performance, system reliability, instrumentation and control

2020 ◽  
Vol 6 ◽  
pp. 43
Author(s):  
Andreas Schumm ◽  
Madalina Rabung ◽  
Gregory Marque ◽  
Jary Hamalainen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Software Systems ◽  
Reactor Performance ◽  
Horizon 2020 ◽  
Term Operation ◽  
In The Beginning ◽  
Electrical Cables

We present a cross-cutting review of three on-going Horizon 2020 projects (ADVISE, NOMAD, TEAM CABLES) and one already finished FP7 project (HARMONICS), which address the reliability of safety-relevant components and systems in nuclear power plants, with a scope ranging from the pressure vessel and primary loop to safety-critical software systems and electrical cables. The paper discusses scientific challenges faced in the beginning and achievements made throughout the projects, including the industrial impact and lessons learned. Two particular aspects highlighted concern the way the projects sought contact with end users, and the balance between industrial and academic partners. The paper concludes with an outlook on follow-up issues related to the long term operation of nuclear power plants.

Thermal Ageing Embrittlement of Casting Duplex Stainless Steels for Nuclear Power Plant

2010 ◽  
Author(s):  
Yuhong Yao ◽  
Jianfeng Wei ◽  
Jiangnan Liu ◽  
Zhengpin Wang ◽  
Yu Wang
Keyword(s):  
Stainless Steels ◽  
Thermal Aging ◽  
Nuclear Power ◽  
Power Plants ◽  
Ageing Time ◽  
Thermal Ageing ◽  
Duplex Stainless Steels ◽  
Impact Properties ◽  
Α Phase ◽  
The Impact

Cast duplex stainless steels (CSS) used for PWR pipes are degraded due to thermal ageing embrittlement during long-term service at 288 °C to 327 °C. Z3CN20-09M Cast duplex Stainless Steels (CSS) made in France for domestic nuclear power plants were thermally aged at 400 °C for 100 h, 300 h, 1000 h, 3000 h and 10000 h. The tensile properties and the impact properties at different thermal aging duration were measured and the effects of the thermal aging time on the microscopic structures and substructures of Z3CN20-09M were respectively investigated by optical microscopy and transmission electron microscopy. The results showed that the tensile strengths of Z3CN20-09M CSS increased gradually with the increment of the thermal ageing time, whereas the impact properties decreased with the prolonging of the thermal ageing time. After long thermal ageing time the dislocation configurations were greatly changed in austenite, and there were precipitates along the austenite-ferrite interface. Moreover, the iron-rich α phase and the chromium-rich α phase precipitated in ferrite aged for 10000h by nucleation and growth rather than the spinodal decomposition. All of above revealed that Z3CN20-09M CSS became brittle during thermal ageing.

Post-Fukushima Research and Development Strategies and Priorities for Water Cooled Reactor Technology Development

2016 ◽  
Author(s):  
Katsumi Yamada ◽  
Abdallah Amri ◽  
Lyndon Bevington ◽  
Pal Vincze
Keyword(s):  
Research And Development ◽  
International Organizations ◽  
Nuclear Power ◽  
Power Plants ◽  
Development Strategies ◽  
Lessons Learned ◽  
Member States ◽  
North East ◽  
Severe Accidents ◽  
Fukushima Daiichi

The Great East Japan Earthquake and the subsequent tsunami on 11 March 2011 initiated accident conditions at several nuclear power plants (NPPs) on the north-east coast of Japan and developed into a severe accident at the Fukushima Daiichi NPP, which highlighted a number of nuclear safety issues. After the Fukushima Daiichi accident, new research and development (R&D) activities have been undertaken by many countries and international organizations relating to severe accidents at NPPs. The IAEA held, in cooperation with the OECD/NEA, the International Experts’ Meeting (IEM) on “Strengthening Research and Development Effectiveness in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant” at IAEA Headquarters in Vienna, Austria, 16–20 February 2015. The objective of the IEM was to facilitate the exchange of information on these R&D activities and to further strengthen international collaboration among Member States and international organizations. One of the main conclusions of the IEM was that the Fukushima Daiichi accident had not identified completely new phenomena to be addressed, but that the existing strategies and priorities for R&D should be reconsidered. Significant R&D activities had been already performed regarding severe accidents of water cooled reactors (WCRs) before the accident, and the information was very useful for predicting and understanding the accident progression. However, the Fukushima Daiichi accident highlighted several challenges that should be addressed by reconsidering R&D strategies and priorities. Following this IEM, the IAEA invited several consultants to IAEA Headquarters, Vienna, Austria, 12–14 May 2015, and held a meeting in order to discuss proposals on possible IAEA activities to facilitate international R&D collaboration in relation to severe accidents and how to effectively disseminate the information obtained at the IEM. The IAEA also held Technical Meeting (TM) on “Post-Fukushima Research and Development Strategies and Priorities” at IAEA Headquarters, Vienna, Austria, 15–18 December 2015. The objective of this meeting was to provide a platform for experts from Member States and international organizations to exchange perspectives and information on strategies and priorities for R&D regarding the Fukushima Daiichi accident and severe accidents in general. The experts discussed R&D topic areas that need further attention and the benefits of possible international cooperation. This paper discusses lessons learned from the Fukushima Daiichi accident based on the presentations and discussions at the meetings mentioned above, and identifies the needs for further R&D activities to develop WCR technologies to cope with Fukushima Daiichi-type accidents.

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