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.

Advanced Inlay System for Inlet/Outlet Nozzles of RV for Preventive Maintenance Against Alloy 600 PWSCC in Japanese PWR Plants

2013 ◽  
Author(s):  
Kazuhide Yamamoto ◽  
Masahiko Kizawa ◽  
Hiroki Kawazoe ◽  
Yuki Kobayashi ◽  
Ken Onishi ◽  
...  
Keyword(s):  
Preventive Maintenance ◽  
Heat Treatments ◽  
Residual Tensile Stress ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Nuclear Plants ◽  
Alloy 690 ◽  
Primary Water

Because many nuclear plants have been in operation for ages, the importance of preventive maintenance technologies is getting higher. One conspicuous problem found in pressurized water reactor (PWR) plants is the primary water stress corrosion cracking (PWSCC) observed in Alloy 600 (a kind of high nickel based alloy) parts. Alloy 600 was used for butt welds between low alloy steel and stainless steel of nozzles of Reactor Vessel (RV), Steam Generator (SG), and Pressurizer (Pz). As PWSCC occurred at these parts may cause Loss of Coolant Accident (LOCA), preventive maintenance is necessary. PWSCC is considered to be caused by a mixture of three elements: high residual tensile stress on surface, material (Alloy 600) and environment. PWSCC can be prevented by improving one of the elements. MHI has been developing stress improvement methods, for example, Water Jet Peening (WJP), Shot Peening by Ultrasonic vibration (USP), and Laser Stress Improvement Process (L-SIP). According to the situation, appropriate method is applied for each part. WJP has been applied for RV nozzles of a lot of plants in Japan. However PWSCC was observed in RV nozzles during the inspection before WJP in recent years, MHI developed the Advanced INLAY system to improve the material from Alloy 600 to Alloy 690. Alloy 600 on the inner surface of the nozzles is removed and welding with Alloy 690 is performed. In addition, heat treatments for the nozzles are difficult for its structural situation, so ambient temperature temper bead welding technique for RV nozzles was developed to make the heat treatments unnecessary. This paper describes the specifications of the advanced INLAY system and introduces the maintenance activities which MHI has applied for three plants in Japan by March 2012.

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.

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.

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.

Operating Procedures: Do they Reduce Operator Errors?

1994 ◽  
Vol 38 (4) ◽  
pp. 205-209 ◽  
Author(s):  
Sharolyn A. Converse
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Error Rates ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Computerized Procedures ◽  
Accident Scenarios ◽  
Accident Conditions ◽  
Subjective Estimates

Computerized operating procedures have been suggested as a mechanism for reducing human error in nuclear power plants. The Computerized Procedures Manual (COPMA-II) is an electronic procedure system that can be used to execute procedures, to track progress through plant procedures, and to automatically monitor plant parameters. To evaluate the effectiveness of COPMA-II, eight teams of two licensed reactor operators operated a scaled pressurized water reactor under normal and accident conditions, using both COPMA-II and traditional paper procedures. Error rates, times to initiate procedures, times to complete procedures, and subjective estimates of workload were collected for each scenario. The most interesting finding of the study was that, for one accident scenario, performance with COPMA-II was twice as accurate as performance with paper procedures. However, operators initiated responses to both accident scenarios fastest with paper procedures. Procedure type did not moderate time to complete procedures.

Flaw Evaluation Procedure for Cast Austenitic Stainless-Steel Materials Using A Newly Developed Statistical Thermal Aging Model

2021 ◽  
Author(s):  
Mohammed F Uddin ◽  
Cédric Sallaberry ◽  
Gery Wilkowski
Keyword(s):  
Statistical Model ◽  
Failure Modes ◽  
Austenitic Stainless Steels ◽  
Limit Load ◽  
Evaluation Procedure ◽  
Chemical Compositions ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Reactor Operating

Abstract Thermal embrittlement of some cast austenitic stainless steels (CASS) occurs at reactor operating temperatures leading to very low fracture toughness. Because of their low aged toughness with high variability, flaw evaluations of CASS material need to be established with an understanding of the materials aged condition, especially since most US Pressurized Water Reactor (PWR) nuclear plants have been given plant life extensions for 60-year operation. A flaw evaluation procedure for CASS materials is presented here using a new statistical model developed to predict the toughness of fully aged CASS using the materials' chemical compositions. In this procedure, the Dimensionless-Plastic-Zone-Parameter (DPZP) analysis is used to determine when limit-load is applicable and also approximate the elastic-plastic correction factor (Z-factor) to predict the failure stress for CASS pipe/fittings with a circumferential surface crack. The procedure was validated against several CF8m pipe test results which include various pipe diameters, crack sizes, ferrite contents, failure modes. The as-developed flaw evaluation procedure was also used to determine the Z-factors for four different pipe diameters for a database of 274 pipe/elbows in US PWR plants -solving 1096 sample problems to understand what range of Z-factors in US PWR plants (for CF8m CASS materials). Finally, the applicability of the CF8m-based statistical model for use with CF3 and CF8 CASS materials was also verified with available test results. This work has been accepted as Code Case N-906 in ASME Boiler and Pressure Vessel (BPV) Code.

The Necessity to Reexamine Key Component Material With Spectrometry Analysis in the Nuclear Island

2019 ◽  
Vol 5 (2) ◽  
Author(s):  
Chongzhi Wu ◽  
Jindong Zhang ◽  
Zongkui Wang ◽  
Zengkui Zhang ◽  
Shenqing Fu
Keyword(s):  
Nuclear Power ◽  
Human Error ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Spectrometry Analysis ◽  
Metal Material ◽  
Component Material

In order to avoid the misuse of metal material in nuclear projects, typical cases happened in advanced passive pressurized water reactor (AP1000) nuclear power projects in China are analyzed. The analysis finding indicates that some cases were caused by defective procedures or undemanding processes performance, and all cases are found to be relevant to human error. It is considered that procedural management cannot completely avoid the misuse of metal material when it is caused by human error, and spectrometry analysis is suggested to reexamine the material of key components.

Leaching Studies of Cement-Glass Package Containing Sodium Borate

1986 ◽  
Vol 84 ◽  
Author(s):  
Masahiro Okamoto ◽  
Koichi Chino ◽  
Tsutomu Baba ◽  
Tatsuo Izumida ◽  
Fumio Kawamura ◽  
...  
Keyword(s):  
Sodium Silicate ◽  
Liquid Waste ◽  
Sodium Borate ◽  
Pressurized Water Reactor ◽  
X Ray Diffraction ◽  
Adsorption Ability ◽  
Pressurized Water ◽  
X Ray ◽  
Solidification Reaction ◽  
Main Components

AbstractA new solidification technique using cement-glass, which is a mixture of sodium silicate, cement, additives, and initiator of the solidification reaction, was developed for sodium borate liquid waste generated from pressurized water reactor (PWR) plants. The cement-glass could solidify eight times as much sodium borate as cement could, because the solidifying reaction of the cement-glass is not hindered by borate ions.The reaction mechanism of sodium silicate and phosphoric silicate (initiator), the main components of cement-glass, was studied through X-ray diffraction and compressive strength measurements. It was found that three- dimensionally bonded silicon dioxide was produced by polymerization of the two silicates. The leaching ratio of cesium from the cement-glass package was one-tenth that of the cement one. This low value was attributed to a high cesium adsorption ability of the cement-glass and it could be theoretically predicted accordingly.

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