Optimization of Maintenance Strategy for Sea Water Pumps in Nuclear Plants

2021 ◽  
Author(s):  
Ling Zhao ◽  
Deyi Liu ◽  
Ming Zhao
Keyword(s):  
Nuclear Power ◽  
Sea Water ◽  
Safety Analysis ◽  
Calculation Model ◽  
Water Pump ◽  
Initial Event ◽  
Maintenance Program ◽  
Nuclear Plants ◽  
Water Pumps ◽  
Service Inspection

Abstract Probabilistic Safety Analysis (PSA) methodology is a significant supplement to the deterministic safety analysis in the nuclear power plant. PSA can be used to evaluate the NPP device change, equipment maintenance, in service inspection. The practicability of modifying sea water pump maintenance programme is evaluated in this paper by determining the initial event probability in fault-tree and modifying the PSA calculation model. Based on the evaluation results, the preventive maintenance of sea water pump and 6kv switch can be changed from executed in the plant overhaul to executed in the routine maintenance. This could make a great contribution in optimizing the NPP sea water pump maintenance program.

Nuclear PWR Plants: Managing the Expertises of Cast Components Removed From Reactor Coolant Circuit

2008 ◽  
Author(s):  
Georges Bezdikian
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Coolant Circuit ◽  
Integrity Assessment ◽  
Reactor Coolant ◽  
Maintenance Program ◽  
Nuclear Plants ◽  
Toughness Evaluation ◽  
Circuit Components ◽  
Prediction Criteria

The French utility has organized the aging degradation assessment for nuclear plants in operation in maximum safety rules. Also the utility decided to engaged large program of expertises, for each degradations identified, based on research and development activities and prediction criteria program of Nuclear Plants in function of several actions: expertises, data bases on characteristics, etc. This paper shows the expertise results and thermal aging for cast duplex stainless steel studies on reactor coolant circuit components engaged to evaluate and to monitor the toughness evaluation and increasing factor. Also this paper presents the strategy associated. This paper shows the applications were evolved for 3-loop PWR plants. The monitoring was mainly oriented on evaluation of the ratio computation and measurement. This integrity assessment and expertises results values available periodically were performed considering: • the life evaluation of the plants and alternative maintenance actions, • the large database from cast reactor coolant component assessed after removed from nuclear power plants, • the identification of degradation for different components and prediction criteria proposed. The results obtained are updated in the periodic maintenance program and in volume of expertise database for life management.

scholarly journals Shaft Failure Sea Water Pumps At the Steam Power Plant Units

2019 ◽  
Vol 9 (1) ◽  
pp. 7008-7012
Keyword(s):  
Power Plant ◽  
Sea Water ◽  
Visual Observation ◽  
Composition Analysis ◽  
Water Pump ◽  
Steam Power ◽  
Steam Power Plant ◽  
Pump Shaft ◽  
Aisi 316 ◽  
Water Pumps

The water pump used to suck and drain seawater to the heat exchanger unit at a Steam Power Plant (PLTU), is damaged. To find out the causal factors of this damaging phenomenon, the fractured shaft is tested which includes visual observation, fractography testing, metallography, hardness testing and chemical composition analysis on a fractured shaft. By knowing the type and cause of damage to the water pump shaft, steps of prevention or prevention can be formulated so that the same damage can be avoided. From the test results, it was found that the average carbon content was lower than the AISI 316 standard. While the average hardness was lower than the standard hardness. The damage that occurs to the CWP water pump shaft is basically caused by fatigue fracture due to excessive workload in the form of dynamic loading

Bayesian Analysis of Piping Failure Frequency Using OECD/NEA Data

2009 ◽  
Author(s):  
M. Wang ◽  
M. D. Pandey ◽  
J. Riznic
Keyword(s):  
Failure Rate ◽  
Nuclear Power ◽  
Probabilistic Models ◽  
Bayesian Procedure ◽  
Two Stage ◽  
Failure Frequency ◽  
Nuclear Plants ◽  
Consistent Manner ◽  
Construction Operation ◽  
Service Inspection

The estimation of piping failure frequency is an important task to support the probabilistic risk analysis and risk-informed in-service inspection of nuclear power plant systems (NPPs). Although various probabilistic models have been proposed in the literature, this paper describes a hierarchical or two-stage Poisson-gamma Bayesian procedure to analyze this problem. In the first stage, a generic distribution of failure rate is developed based on the failure observations from a group of similar plants. This distribution represents the interplant (plant-to-plant) variability arising from differences in construction, operation and maintenance conditions. In the second stage, the generic prior obtained from the first stage is updated by using the data specific to a particular plant, and thus a posterior distribution of plan specific failure rate is derived. The two-stage Bayesian procedure is able to incorporate different levels of variability in a more consistent manner. The proposed approach is applied to estimate the failure frequency using the OECD/NEA pipe leakage data for the U.S. nuclear plants.

In-Service Inspection and On-Line Monitoring of Inaccessible Components in Nuclear Power Plants Using Guided Wave Technology

2014 ◽  
Author(s):  
Florin Turcu ◽  
Mauro Cappelli ◽  
Davide Mazzini ◽  
Sergio Pistelli ◽  
Marco Raugi
Keyword(s):  
Nuclear Power ◽  
Scientific Community ◽  
Power Plants ◽  
Guided Waves ◽  
Guided Wave ◽  
Ultrasonic Guided Waves ◽  
Critical Parameters ◽  
Nuclear Plants ◽  
On Line ◽  
Service Inspection

One of the most challenging problems in the on-line monitoring of critical parameters of nuclear plants is the inspection of components that result inaccessible or difficult to reach. In this context, there is an increasing interest of the scientific community and industry for the use of Ultrasonic Guided Waves (UGW) for addressing this issue. In this work, the problem of the applicability of the UGW technique with magnetostrictive sensors to NPP structures is described, together with the outline of the related advantages as well as the main technical concerns that may arise from such applications. This methodology has been tested on experimental activities concerning high temperature applications. Results show the effectiveness of such an approach.

Study on Typical Initial Event of ATWS for Small Modular Reactor

2016 ◽  
Author(s):  
Zhang Dan ◽  
Ran Xu ◽  
Qiu Zhifang ◽  
Zhou Ke ◽  
Feng Li
Keyword(s):  
Optimum Design ◽  
Safety Analysis ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Reactor Coolant ◽  
Initial Event ◽  
Small Modular Reactor ◽  
Activation System

The method for ATWS (anticipated transient without scram) analysis was completely developed for commercial pressurized water (PWR) reactor plants, especially for selecting of typical initial events. For accident analysis of ATWS, it is different between PWR and small modular reactor (SMR), as different structures and characters, and it is necessary to study the typical initial events for these reactors. Based on the standard of PWR, the demanding for ATWS analysis was studied and the consequences for typical anticipated transient was calculated using RELAP5/MOD3.2 code, “maintain reactor coolant pressure boundary integrity” was selected as limiting criterion. The results shows for SMR, anticipated transient with the most serious consequence for ATWS are loss of offsite power and inadvertent control rod withdraw event, this conclusion will support to prepare the safety analysis report and optimum design of diversity activation system (DAS) for SMR.

Architecting “Maintenance Program” for Piping and Pipe Support

2013 ◽  
Author(s):  
Koichi Tsumori ◽  
Yoshizumi Fukuhara ◽  
Hiroyuki Terunuma ◽  
Koji Yamamoto ◽  
Satoshi Momiyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Maintenance Management ◽  
3D Cad ◽  
Cad System ◽  
Nuclear Disaster ◽  
Safety Regulations ◽  
Maintenance Program

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations. The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.

Wave Energy-Driven Resonant Sea-Water Pump

1997 ◽  
Vol 119 (3) ◽  
pp. 191-195 ◽  
Author(s):  
S. P. R. Czitrom
Keyword(s):  
Numerical Model ◽  
Sea Water ◽  
Scale Model ◽  
Water Pump ◽  
Incoming Wave ◽  
Frequency Wave ◽  
Variable Volume ◽  
Mass Spring ◽  
Air Compression ◽  
Compression Chamber

A wave-driven sea-water pump which operates by resonance is described. Oscillations in the resonant and exhaust ducts perform similar to two mass-spring systems coupled by a third spring acting for the compression chamber. Performance of the pump is optimized by means of a variable volume air compression chamber (patents pending) which tunes the system to the incoming wave frequency. Wave tank experiments with an instrumented, 1:20-scale model of the pump are described. Performance was studied under various wave and tuning conditions and compared to a numerical model which was found to describe the system accurately. Successful sea trials at an energetic coastline provide evidence of the system’s viability under demanding conditions.

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