The Root Cause Analysis of the Bolt Fracture in the Nuclear Power Plant Essential Service Water System

2017 ◽  
Author(s):  
Li Nan ◽  
Lao Yi ◽  
Che Yinhui
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Water System ◽  
Element Distribution ◽  
Material Microstructure ◽  
Root Cause ◽  
Check Calculation ◽  
Service Water ◽  
Essential Service

When inspecting in the nuclear power plant, the bolt of the 001/004 pump in Essential Service Water system was found fracture. The bolt in 001 pump had ever fractured before, and it had been replaced. In this paper, the material, microstructure, energy dispersive spectrometry and mechanical check calculation of the bolt are analyzed. The result shows, the bolt breakage is for stress corrosion cracking, the corrosion element is Cl−. When the martensitic stainless steel is in the heat treatment, the temperature is improper control, which causing the Cr element distribution changed. So the ability of material to resist corrosion becomes poor which is the root cause of the bolt fracture.

Research on Operating Strategy of Total Loss of Essential Service Water System With RHR Not Connected in PWR Unit

2020 ◽  
Author(s):  
Shengtao Zhang ◽  
Ke Yi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Sink ◽  
Nuclear Power ◽  
Cooling System ◽  
Water System ◽  
Total Loss ◽  
Operating Strategy ◽  
Service Water ◽  
Essential Service

Abstract Essential Service Water System (WES) is part of the nuclear power plant cooling system which provides the final heat sink for nuclear power plants. Therefore, WES must operate stably, safely and reliably for a long time. The total loss of WES accident is a design extended condition and will result in the loss of the final heat sink of the unit. The consequences of the accident are severe. In order to deal with the accident quickly and effectively and ensure the safety and economics of the power plant in accident condition, it’s necessary to formulate corresponding treatment strategy to deal with the transient. This paper developed a strategy for dealing with the total loss of WES with Residual Heat Removal System (RHR) not connected condition in Generation III nuclear power plant. The structure of the WES system and the types of failures that may occur are analyzed, and thus the symptoms of the faults are obtained and the entry conditions for the operating strategy are determined. The effect of faults on unit equipment and safety functions and the impact on nuclear steam supply system (NSSS) control are analyzed in this paper. Combined with the unit design, the system and equipment for controlling and mitigating related safety functions are analyzed, and the mitigation method and the fallback strategy of the fault are determined. Thereby a complete operating strategy of total loss of WES with RHR not connected is obtained. In addition, this paper analyzes and evaluates the operating strategy by simulating thermal hydraulic calculation for the first time. The results show that without staff intervention Component Cooling System (WCC) temperature reached 55°C limits after running a few minutes. Based on the intervention of the operating strategy proposed in this paper, WCC temperature reached the 55°C limits when the unit was operated at about 4 hours and 55 minutes. The result shows that and the strategy can effectively alleviate the failure and provide sufficient intervention time for the operator to bring the unit to a safe state.

Investigation of Effects of Piping Configuration and Water Supply Pressure on Air Intrusion

2012 ◽  
Author(s):  
Jaehyok Lim ◽  
Basar Ozar ◽  
Christopher E. Henry ◽  
Kevin B. Ramsden
Keyword(s):  
Power Plant ◽  
Water Supply ◽  
Storage Tank ◽  
Nuclear Power ◽  
Water Pressure ◽  
Water System ◽  
Supply Pressure ◽  
Service Water ◽  
Essential Service ◽  
Supply Water

An evaluation of the effects of geometry and water supply pressure on the void transport has been performed using RELAP5/MOD3.3 (patch03). Two different piping configurations were considered for a hypothetical nuclear power plant. The cases that were analyzed considered switchover between two different water supplies, i.e. condensate storage tank (CST) and essential service water system (SX) for a safety system that acted as the ultimate heat sink. In addition, two different pressures were considered for the pressure of SX to investigate the effect of supply water pressure on void transport. Results were interpreted based on the differences in the geometries of the piping configurations and supply water pressures.

Root Cause of Thermal Sleeve Loosening in a Korean Standard Nuclear Power Plant

2006 ◽  
Author(s):  
Sang-Nyung Kim ◽  
Sang-Gyu Lim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Measurement Data ◽  
Significant Loss ◽  
Design Modification ◽  
Root Cause ◽  
Explosive Expansion ◽  
The Impact

The safety injection (SI) nozzle of a 1000MWe-class Korean standard nuclear power plant (KSNP) is fitted with thermal sleeves (T/S) to alleviate thermal fatigue. Thermal sleeves in KSNP #3 & #4 in Yeonggwang (YG) & Ulchin (UC) are manufactured out of In-600 and fitted solidly without any problem, whereas KSNP #5 & #6 in the same nuclear power plants, also fitted with thermal sleeves made of In-690 for increased corrosion resistance, experienced a loosening of thermal sleeves in all reactors except KSNP YG #5-1A, resulting in significant loss of generation availability. An investigation into the cause of the loosening of the thermal sleeves only found out that the thermal sleeves were subject to severe vibration and rotation, failing to uncover the root cause and mechanism of the loosening. In an effort to identify the root cause of T/S loosening, three suspected causes were analyzed: (1) the impact force of flow on the T/S when the safety SI nozzle was in operation, (2) the differences between In-600 and In-690 in terms of physical and chemical properties (notably the thermal expansion coefficient), and (3) the positioning error after explosive expansion of the T/S as well as the asymmetric expansion of T/S. It was confirmed that none of the three suspected causes could be considered as the root cause. However, after reviewing design changes applied to the Palo Verde nuclear plant predating KSNP YG #3 & #4 to KSNP #5 & #6, it was realized that the second design modification (in terms of groove depth & material) had required an additional explosive energy by 150% in aggregate, but the amount of gunpowder and the explosive expansion method were the same as before, resulting in insufficient explosive force that led to poor thermal sleeve expansion. T/S measurement data and rubbing copies also support this conclusion. In addition, it is our judgment that the acceptance criteria applicable to T/S fitting was not strict enough, failing to single out thermal sleeves that were not expanded sufficiently. Furthermore, the T/S loosening was also attributable to lenient quality control before and after fitting the T/S that resulted in significant uncertainty. Lastly, in a flow-induced vibration test planned to account for the flow mechanism that had a direct impact upon the loosening of the thermal sleeves that were not fitted completely, it was discovered that the T/S loosening was attributable to RCS main flow. In addition, it was proven theoretically that the rotation of the T/S was induced by vibration.

Based on Human Behavior Process of Human Error Defensive Management Research for NPP

2017 ◽  
Author(s):  
Shen Yang ◽  
Geng Bo ◽  
Li Dan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Control Chart ◽  
Human Behavior ◽  
Nuclear Power ◽  
Human Error ◽  
Human Performance ◽  
Human Factor ◽  
Error Management ◽  
Root Cause

According to the research of nuclear power plant human error management, it is found that the traditional human error management are mainly based on the result of human behavior, the event as the point cut of management, there are some drawbacks. In this paper, based on the concept of the human performance management, establish the defensive human error management model, the innovation point is human behavior as the point cut, to reduce the human errors and accomplish a nip in the bud. Based on the model, on the one hand, combined with observation and coach card, to strengthen the human behavior standards expected while acquiring structured behavior data from the nuclear power plant production process; on the other hand, combined with root cause analysis method, obtained structured behavior data from the human factor event, thus forming a human behavior database that show the human performance state picture. According to the data of human behavior, by taking quantitative trending analysis method, the P control chart of observation item and the C control chart of human factor event is set up by Shewhart control chart, to achieve real-time monitoring of the process and result of behavior. At the same time, development Key Performance Indicators timely detection of the worsening trend of human behavior and organizational management. For the human behavior deviation and management issues, carry out the root cause analysis, to take appropriate corrective action or management improvement measures, so as to realize the defense of human error, reduce human factor event probability and improve the performance level of nuclear power plant.

Residual Stress Distributions and Guidelines for Crack Inspection in Pipe Bends

2012 ◽  
Author(s):  
Metin Yetisir ◽  
Zane Walker
Keyword(s):  
Residual Stress ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Significant Role ◽  
Nuclear Power ◽  
Stress Distributions ◽  
Pipe Bend ◽  
Comprehensive Review ◽  
Stress Measurements ◽  
Root Cause

Root cause investigations of feeder pipe cracks detected in one CANDU™ nuclear power plant indicated that the residual stress had a significant role in these failures. As a result, numerous residual stress measurements of pipe bends were obtained and models for predicting the residual stress distributions associated with various bending processes were developed. This paper provides a comprehensive review of pipe bend residual stress data and provides guidelines for identifying the most susceptible locations for targeted crack inspections. Residual stress data, generated since 1997 as part of the CANDU feeder cracking investigation, was compiled and presented for the quick dissemination of information. This information is summarized in quick lookup tables where likely crack locations are identified for pipe bends manufactured using various fabrication techniques.

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