The Analysis of Initiating Events in PSA for TMSR

2016 ◽  
Author(s):  
Jiaxu Zuo ◽  
Jianping Jing ◽  
Wei Song ◽  
Chunming Zhang
Keyword(s):  
Operating Experience ◽  
Safety Analysis ◽  
Nuclear Safety ◽  
Analysis Method ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Analysis And Evaluation ◽  
Key Points ◽  
Analysis Calculation ◽  
High Temperature Reactor

The initiation events analysis and evaluation were the beginning of nuclear safety analysis and probabilistic safety assessment, and it was the key points of the nuclear safety analysis. The main methods of the initiation events analysis are reference to existing lists and reports, operating experience, project evaluation and logic diagrams analysis. Currently, the initiation events analysis method and experiences both focused on pressurized water reactor but there are no general theories for Fluoride Salt-Cooled High Temperature Reactor (FHR). With FHR’s research and development, the initiation events analysis and evaluation was increasingly important. Based on the FHR’s design, the theories and methods of initiation events analysis would be researched and developed. From the FHR’s design, the systems, subsystems and components are divided to identify the safety functions of them. Base on the safety functions, the logical analysis and accident analysis calculation method would be combined to study FHR’s initiation events. The theory of analysis would be developed and the analysis method system would be discussed. Finally, the preliminary initiation events list of FHR will be discussed and researched. The results would help TMSR’s reactor designs and nuclear safety analysis.

Use of Risk Insights to Enhance Safety Focus of Small Modular Reactor Reviews

2011 ◽  
Author(s):  
Stewart L. Magruder
Keyword(s):  
Review Process ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water Reactor ◽  
Design Features ◽  
Pressurized Water ◽  
Analysis And Evaluation ◽  
Small Modular Reactor ◽  
Efficiency And Effectiveness ◽  
Regulatory Commission ◽  
Regulatory Controls

The U.S. Nuclear Regulatory Commission staff plans to apply a more integrated, graded approach to the review of small modular reactor (SMR) pre-application activities and design applications. The concept is to improve the efficiency and effectiveness of the reviews by focusing on safety significant structures, systems, and components (SSCs). The unique design features associated with SMRs and knowledge gained reviewing other passive reactor designs present opportunities to risk-inform the SMR design certification process to a greater extent than previously employed. The review process can be modified for SMR applications by considering the aggregate of regulatory controls pertaining to SSCs as part of the review and determining those regulatory controls which may supplement or replace, as appropriate, part of the technical or engineering analysis and evaluation. Risk insights acquired from staff reviews of passive LWR designs (i.e. AP1000, ESBWR) can also be incorporated into the review process. Further, risk insights associated with integral pressurized water reactor (iPWR) design features (i.e. Underground facilities impact on turbine missiles review) can be incorporated into the review process.

Risk-Informed Preventive Maintenance Optimization

2004 ◽  
Author(s):  
James K. Liming ◽  
James E. Salter
Keyword(s):  
Asset Management ◽  
Preventive Maintenance ◽  
Human Error ◽  
Failure Modes ◽  
Cost Benefit ◽  
Nuclear Safety ◽  
Target System ◽  
Maintenance Optimization ◽  
Pressurized Water Reactor ◽  
Pressurized Water

Past and ongoing electric generating station owner investments in plant information technology (such as database query applications and other client workstation tools) have made it possible for plant staffs to utilize information contained in the work management systems to quickly link equipment failure modes to related preventative maintenance (PM) activities. A typical pressurized water reactor feedwater (FW) system is applied as the “target system” for examples in this paper. This typical FW system is comprised of approximately 3,800 “tag” or “part number” items which in turn represent about 16,300 failure modes. Effective risk-informed asset management (RIAM) of FW preventive maintenance (PM) activities requires these failure modes to be modeled in a plant availability model. In this paper we present development of a process for supporting PM optimization, applying cost-benefit-risk analysis and RIAM tools and techniques. In this preventive maintenance optimization (PMO) process, PM activities are evaluated for their projected impacts on plant profitability and nuclear safety. PM activities (PMs) are “optimized” for desirable impact to help ensure electric utilities maintain or improve upon high levels of nuclear safety and profitability. In this PMO application the level of detail of the target system(s) is enhanced to support plant decision-making at the component failure mode and human error mode level of indenture. Results of case studies in FW system PMO using typical plant data are presented.

Utilization of NEK full scope simulator for plant operation optimization, nuclear education and engineering in 20 years

2021 ◽  
Author(s):  
Matjaž Žvar ◽  
Tomaž Žagar
Keyword(s):  
Emergency Preparedness ◽  
Nuclear Power ◽  
Impact Analysis ◽  
Training Session ◽  
Nuclear Safety ◽  
Severe Accident ◽  
Electricity Production ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water

Abstract This paper gives an impact analysis of utilization of NPP full scope simulator on operation parameters, training and education in nuclear power plant Krško. The Slovenian Nuclear Safety Administration issued their simulator decree to NEK in April 1995. The first training session on the simulator was performed in April 17th 2000 and since then the simulator has been used on daily bases to improve operator knowledges, skills and performances. At the time, this was the first full scope simulator with the capability to simulate Beyond design basis accidents (severe accidents). The ability to simulate core meltdown and containment breach made it very suitable for emergency preparedness drills. After the 2017 simulator upgrade, fuel meltdown in the spent fuel pool can be simulated using the Modular Accident Analysis Program – MAAP5. This capability is still unique for full scope simulators even today. The simulator is also used for pre-testing of plant modifications before their implementation on site or for just-in-time training for infrequent performed evolutions or for procedure development and testing. The Pressurized Water Reactor Owners Group (PWROG) used the NEK simulator in 2018 to develop the new set of the Severe Accident Management Guidelines, incorporated with a completely new usage approach. In all of these years, the simulator has been actively participating in the increased reliability and stability of the electricity production and in achieving NEK's vision to be a worldwide leader in nuclear safety and excellence.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Line Down Horizontal Elbow

2020 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Thermal Transients ◽  
Reactor Coolant ◽  
Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from various thermal transients such as the reactor coolant system (RCS) sampling and excess letdown may also contribute to the failure of RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Section III Class 1 piping stress formula, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Neutron Embrittlement Analysis of Qinshan 320MWe RPV on OLE

2017 ◽  
Author(s):  
Tao Hongxin ◽  
He Yinbiao ◽  
Cao Ming ◽  
Shen Rui
Keyword(s):  
Service Life ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Nuclear Safety ◽  
Radioactive Material ◽  
Pressurized Water Reactor ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Key Factor

One of the fundamental requirements on nuclear safety is to prevent the radioactive material from being released. Therefore, it is paramount to maintain the structural integrity of the pressure boundary of the reactor coolant system. The reactor pressure vessel (RPV), under high temperature, high pressure and high radiation in operation, is the most important as well as a Class I nuclear safety equipment. For a pressurized water reactor (PWR), the life of the RPV determines the service life of the entire nuclear power plant. The key factor controlling the life of a RPV is the accumulation of the neutron flux and which induces irradiation embrittlement degrading the anti-fracture capability of the RPV material. Several anti-fracture capability assessments carried out for the Qinshan 320MWe (QS1) RPV, such as (a) the structural integrity assessment against pressurized thermal shocks; (b) the fracture mechanics assessment under irradiation; (c) the P-T limit curves revised; (d) the evaluation of USE. They all demonstrated that the structural integrity of the QS1 RPV would be maintained for the extended service life.

The safety analysis of a small pressurized water reactor utilizing fully ceramic microencapsulated fuel

2017 ◽  
Vol 320 ◽  
pp. 250-257 ◽  
Author(s):  
Jinfeng Huang ◽  
Ning Li ◽  
Yaoli Zhang ◽  
Qixun Guo ◽  
Jian Zhang
Keyword(s):  
Safety Analysis ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

scholarly journals Research On Optimization Method of Fatigue Analysis for Nuclear Pipeline

2021 ◽  
Vol 257 ◽  
pp. 02017
Author(s):  
Yuan Liang ◽  
Liu Shengyong ◽  
Yang Jie ◽  
Zhou Qiang ◽  
Zhang Guihe
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Optimization Method ◽  
Fatigue Analysis ◽  
Analysis Method ◽  
Pressurized Water Reactor ◽  
Original Design ◽  
Pressurized Water ◽  
Design Life ◽  
License Application

The initial design life of nuclear power plant is 40 years. In 60 year life extending license application, the fatigue of component should be evaluated under the influence of the fatigue factors of the pressurized water reactor coolant environment. Because the original design used a more conservative analysis method, the result could not meet the requirement of Cumulative usage fatigue factor of RCC-M. An optimizing analysis method is studied, and as an example of application, optimizing fatigue analysis of Safety Injection Nozzle of Main Coolant Line is performed. The evaluation results show that the optimized fatigue analysis results meet the requirements of RCC-M.

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