Evaluation of the Integrity of Reactor Pressure Vessels in Nuclear Power Plants during Pressurized Thermal Shock Events

2017 ◽  
Vol 2017 (0) ◽  
pp. J0310203
Author(s):  
Xiaoyong RUAN ◽  
Toshiki NAKASUJI ◽  
Kazunori MORISHITA
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

Proposal for Update on Evaluation Procedure for Reactor Pressure Vessels Against Pressurized Thermal Shock Events in Japan

2014 ◽  
Author(s):  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto ◽  
Naoki Ogawa
Keyword(s):  
Fracture Toughness ◽  
Thermal Shock ◽  
Nuclear Power ◽  
Pressure Vessels ◽  
Evaluation Procedure ◽  
Pressurized Thermal Shock ◽  
Vessel Integrity ◽  
Inspection Techniques ◽  
Reactor Pressure ◽  
Current Procedure

The evaluation procedure for the reactor pressure vessel integrity of Japanese PWR plants against Pressurized Thermal Shock (PTS) events is prescribed in the Japan Electric Association Code, JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” since 1991. The current procedure was developed based on the PTS verification test program, which was conducted as Japanese national project and the related studies in 1980’s. Since much progress has been made on fracture mechanics, fracture toughness, in-service inspection techniques/results and so on, it is preferred to advance the current procedure for more credible evaluation by reflecting the latest knowledge. This paper describes the outline of the studies to update the current procedure.

Probabilistic Structural Integrity Analysis of Reactor Pressure Vessels During Pressurized Thermal Shock Events

2013 ◽  
Vol 136 (1) ◽  
Author(s):  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Kunio Onizawa
Keyword(s):  
Crack Initiation ◽  
Thermal Shock ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Sensitivity Analyses ◽  
Conditional Probabilities ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

To apply a probabilistic fracture mechanics (PFM) analysis to the structural integrity assessment of a reactor pressure vessel (RPV), a PFM analysis code has been developed at JAEA. Using this PFM analysis code, pascal version 3, the conditional probabilities of crack initiation (CPIs) and fracture for an RPV during pressurized thermal shock (PTS) events have been analyzed. Sensitivity analyses on certain input parameters were performed to clarify their effect on the conditional fracture probability. Comparisons between the conditional probabilities and the temperature margin (ΔTm) based on the current deterministic analysis method were made for various model plant conditions for typical domestic older types of RPVs. From the analyses, a good correlation between ΔTm and the conditional probability of crack initiation was obtained.

Upgrading the Operational Safety of Nuclear Power Plants Through the TACIS Nuclear Safety Assistance Programme

2004 ◽  
Author(s):  
M. Bie`th ◽  
R. Ahlstrand ◽  
C. Rieg ◽  
P. Trampus
Keyword(s):  
European Commission ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Reactor Pressure Vessel ◽  
Nuclear Safeguards ◽  
Integrity Assessment ◽  
Reactor Pressure

The European Union’ TACIS programme was established for the New Independent States since 1991. One priority for TACIS funding is nuclear safety. The European Commission has made available a total of € 944 million for nuclear safety programmes covering the period 1991–2003. The TACIS nuclear safety programme is devoted to the improvement of the safety of Soviet designed nuclear installations in providing technology and safety culture transfer. The Joint Research Center (JRC) of the European Commission is carrying out works in the following areas: • On-Site Assistance for TACIS Nuclear Power Plants; • Design Safety and Dissemination of TACIS results; • Reactor Pressure Vessel Embrittlement for VVER in Russia and Ukraine; • Regulatory Assistance; • Industrial Waste Management and Nuclear Safeguards. This paper gives an overview of the Scientific and Technical support that JRC is providing for the programming and the implementation of the TACIS nuclear safety programmes. In particular, two new projects are being implemented to get an extensive understanding of the VVER reactor pressure vessel embritttlement and integrity assessment.

The Position Design for M310 Reactor Vessel’s Instrumentation Penetration Based on None-Alignment Analysis

2013 ◽  
Author(s):  
Xin Xia ◽  
Hua Du ◽  
Ning Li ◽  
Bin Xu ◽  
Yan Li ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Economic Losses ◽  
Design Value ◽  
Alignment Analysis ◽  
Reactor Pressure ◽  
Reactor Internals

The position of the reactor pressure vessel’s instrumentation penetration is the most important factor that influences the none-alignment between the instrumentation penetration and the instrumentation tube of reactor internals. The more the none-alignment, the more wear the flux thimble will suffer, which will damage the flux detector and cause economic losses. This paper analyzes the none-alignment between the instrumentation penetration and the instrumentation tube, and combines the feedback of the experience in nuclear power plants having been built or being build, puts forward the reasonable design value for the instrumentation penetration’s position.

Background on Low Temperature Overpressure Protection System Setpoints for Pressure-Temperature Curves

2020 ◽  
Author(s):  
Jeffrey C. Poehler ◽  
Gary L. Stevens ◽  
Anees A. Udyawar ◽  
Amy Freed
Keyword(s):  
Low Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Pressurized Water Reactors ◽  
Documentation System ◽  
Pressurized Water ◽  
System Pressure ◽  
Water Reactors

Abstract ASME Code, Section XI, Nonmandatory Appendix G (ASME-G) provides a methodology for determining pressure and temperature (P-T) limits to prevent non-ductile failure of nuclear reactor pressure vessels (RPVs). Low-Temperature Overpressure Protection (LTOP) refers to systems in nuclear power plants that are designed to prevent inadvertent challenges to the established P-T limits due to operational events such as unexpected mass or temperature additions to the reactor coolant system (RCS). These systems were generally added to commercial nuclear power plants in the 1970s and 1980s to address regulatory concerns related to LTOP events. LTOP systems typically limit the allowable system pressure to below a certain value during plant operation below the LTOP system enabling temperature. Major overpressurization of the RCS, if combined with a critical size crack, could result in a brittle failure of the RPV. Failure of the RPV could make it impossible to provide adequate coolant to the reactor core and result in a major core damage or core melt accident. This issue affected the design and operation of all pressurized water reactors (PWRs). This paper provides a description of an investigation and technical evaluation regarding LTOP setpoints that was performed to review the basis of ASME-G, Paragraph G-2215, “Allowable Pressure,” which includes provisions to address pressure and temperature limitations in the development of P-T curves that incorporate LTOP limits. First, high-level summaries of the LTOP issue and its resolution are provided. LTOP was a significant issue for pressurized water reactors (PWRs) starting in the 1970s, and there are many reports available within the U.S. Nuclear Regulatory Commission’s (NRC’s) documentation system for this topic, including Information Notices, Generic Letters, and NUREGs. Second, a particular aspect of LTOP as related to ASME-G requirements for LTOP is discussed. Lastly, a basis is provided to update Appendix G-2215 to state that LTOP setpoints are based on isothermal (steady-state) conditions. This paper was developed as part of a larger effort to document the technical bases behind ASME-G.

scholarly journals Probabilistic Fracture Mechanics Round Robin Analysis of Reactor Pressure Vessels during Pressurized Thermal Shock

2010 ◽  
Vol 47 (12) ◽  
pp. 1131-1139 ◽  
Author(s):  
Myung Jo JHUNG ◽  
Seok Hun KIM ◽  
Young Hwan CHOI ◽  
Yoon Suk CHANG ◽  
Xiangyuan XU ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Round Robin ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

Structural Integrity for the RPV of French NPP during the LRA

2016 ◽  
Vol 853 ◽  
pp. 453-457
Author(s):  
Ming Ya Chen ◽  
Wei Wei Yu ◽  
Jin Hua Shi ◽  
Rong Shan Wang ◽  
Lv Feng ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
The Other ◽  
Light Water Reactor ◽  
Original Design ◽  
Light Water ◽  
Pressurized Thermal Shock ◽  
The Usa ◽  
Reactor Pressure

Most of the French Nuclear Power Plants (NPPs) are currently embarking upon efforts to renew their operating license, while the pressurized thermal shock (PTS) events and environmentally assisted fatigue (EAF) pose potentially significant challenges to the structural integrity of the reactor pressure vessel (RPV) which has the potential to be NPP life-limiting conditions. In the assessment of the PTS events, the deterministic fracture mechanics (DFM) is still used as the basic mechanics in most countries except for the USA. While the maximum nil-ductility-transition temperature (RTNDT) is about 80°C for 54 French RPVs after 40 years operation, the maximum allowable RTNDT is only about 70 oC and 80 oC for the typical PTS events in the IAEA and NEA reports, respectively. On the other hand, the effects of light water reactor (LWR) environmental (other than moderate environment in the code) were not considered in the original design, while the effects of LWR environmental are needed to be considered in the LRA according to the USA regulations. In this paper, the challenges of the PTS and EAF are discussed, and some suggestions are also given for the LRA

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