scholarly journals Dynamic Reliability Evaluation of Diesel Generator System of One Chinese 1000MWe NPP Considering Temporal Failure Effects

2021 ◽  
Vol 9 ◽  
Author(s):  
Dingqing Guo ◽  
Manjiang Yang ◽  
Hongmei Wu ◽  
Daochuan Ge ◽  
Xuewu Cao
Keyword(s):  
Power Systems ◽  
Nuclear Power ◽  
Markov Chain Model ◽  
Block Diagram ◽  
Fault Tree ◽  
Chain Model ◽  
Diesel Generator ◽  
Generator System ◽  
Pressurized Water ◽  
Dynamic Reliability

Loss of power supply from the diesel generator system (DGS) after loss of offsite power (LOOP) will pose great threat to the safety of GEN-II pressurized water reactors (PWR). Therefore, it is very desirable to evaluate the DGS’s reliability. The traditional analyzing tools are limited to static approaches neglecting the dynamic sequence failure behaviors, such as reliability block diagram (RBD), static fault tree (SFT). Static reliability modeling techniques are not capable of capturing the dynamic sequence-dependent failure behaviors typically existing in NPP safety systems such as DGS, and thus often overestimate the unreliability of systems. In this paper, motivated to study the effects of sequence failure behaviors, dynamic fault tree (DFT) is applied to evaluate the reliability of the DGS of one Chinese 1000MWe Nuclear Power Plant (NPP), and an integrated two-phased Markov Chain model is also developed, which can be considered as a contribution of this article. Comparative study of DGS reliability between DFT and SFT is carried out. The results indicate that compared with the result derived from the DFT model, the unreliability of DGS calculated by SFT is greatly overestimated by about one to two orders of magnitude. Therefore, DFT has a potential to improve the economy of NPP by relaxing the overestimated unreliability of nuclear power systems.

The Fault Tree Analysis on Auxiliary Feedwater Pump Room Ventilation System in Pressurized Water Reactor Power

2013 ◽  
Author(s):  
Yang Li ◽  
Chen Hang
Keyword(s):  
Nuclear Power ◽  
Fault Tree ◽  
Fault Tree Analysis ◽  
Easy Access ◽  
Main Function ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tree Analysis ◽  
Water Reactor ◽  
Mandatory Requirement

Main function of HVAC is to remove heat from equipment and pipeline, hold the inner condition, maintain an ambient temperature and humidity that keep the equipments function properly and easy access. Although regulation is no mandatory requirement of redundant equipment design and preservation function in case of specified disaster or man-made accident. In fact, It does be influenced by the incident whether partial failure or full. The hazard factor determination and qualitative analysis are based on fault tree analysis through simulated mode from selected the typical system. The identification of accident cause, hazard cause and fault mode is essential for improving system reliability. According the analysis result, It will be optimization factor such as installation and design process, maintenance ability, material plan, corrosion preventing. It’s helpful to control hazard under accepted level. This method given in the article is a new way to treat HVAC system in pressurized water reactor nuclear power. It hopes that this method will lead to reduce accident loss, save maintenance fee, bring economic benefits and improve the risk of nuclear power.

Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 164-172
Author(s):  
R. A. Fahmy ◽  
R. I. Gomaa
Keyword(s):  
Power Plant ◽  
System Reliability ◽  
Nuclear Power ◽  
Fault Tree ◽  
Fault Tree Analysis ◽  
Cost Effective ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tree Analysis ◽  
Dynamic Fault Tree

Abstract The safe and secure designs of any nuclear power plant together with its cost-effective operation without accidents are leading the future of nuclear energy. As a result, the Reliability, Availability, Maintainability, and Safety analysis of NPP systems is the main concern for the nuclear industry. But the ability to assure that the safety-related system, structure, and components could meet the safety functions in different events to prevent the reactor core damage requires new reliability analysis methods and techniques. The Fault Tree Analysis (FTA) is one of the most widely used logic and probabilistic techniques in system reliability assessment nowadays. The Dynamic fault tree technique extends the conventional static fault tree (SFT) by considering the time requirements to model and evaluate the nuclear power plant safety systems. Thus this paper focuses on developing a new Dynamic Fault Tree for the Auxiliary Feed-water System (AFWS) in a pressurized water reactor. The proposed dynamic model achieves a more realistic and accurate representation of the AFWS safety analysis by illustrating the complex failure mechanisms including interrelated dependencies and Common Cause Failure (CCF). A Simulation tool is used to simulate the proposed dynamic fault tree model of the AFWS for the quantitative analysis. The more realistic results are useful to establish reliability cantered maintenance program in which the maintenance requirements are determined based on the achievement of system reliability goals in the most cost-effective manner.

Reliability Analysis and Availability Enhancements for Redundant Systems With Reparable Components

2016 ◽  
Author(s):  
Shuqiao Zhou ◽  
Duo Li ◽  
Chao Guo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Markov Chain Model ◽  
Life Time ◽  
Theoretical Models ◽  
Chain Model ◽  
Monitoring And Control ◽  
Monitoring Methods ◽  
Under Construction ◽  
Redundant Systems

Redundancy is widely used for enhancing a system’s overall availability. As an HTR demonstrated plant, a high temperature gas-cooled reactor pebble-bed module (HTR-PM) now is under construction in China and the construction will be completed around 2017. In HTR-PM, there are many devices and device groups used in a redundant way to guarantee the high availability of the related functions, especially the functions shared by two reactors during the entire life time. It is very important and necessary to determine their reliabilities as well as how to make a decision about the related maintenance policies to enhance their availabilities. In this paper, typical redundant styles in the HTR-PM are summarized and demonstrated. Accordingly, the theoretical models, which are able to describe the reliabilities of the redundant systems, are proposed based on Markov chain model. Moreover, for a specific redundant structure, the relationship between the availability and the maintenance period is analytically addressed. Based on the model, we address that: as the digital monitoring and control technologies are widely used in nuclear power plants, monitoring methods targeting for decreasing maintenance costs and meanwhile increasing the availabilities for different redundant styles are very beneficial.

Evaluation of the forced outage length contribution of critical nuclear power systems in an extended operating cycle

2000 ◽  
Vol 214 (1) ◽  
pp. 23-28
Author(s):  
C W Kang
Keyword(s):  
Power Systems ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Nuclear Regulatory Commission ◽  
Capacity Factor ◽  
Plant Performance ◽  
Operating Cycle ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission

The work presented here presents an evaluation method for the question of how reliably the system (or component), responsible for the dominant plant availability loss, will run in an extended 48 month operating cycle. As major contributors to the total plant forced outage time in pressurized water reactors (PWRs), reactor coolant pumps (RCPs) and main feed pumps (MFPs) are chosen as specific example systems for a case study. The method proposed estimates the expected forced outage length contribution of each system to the maximum allowed outage length given a certain plant capacity factor. Based upon the current reliability level estimated from the Nuclear Regulatory Commission plant performance database, the assessment of each system impact shows that 14.2 and 2.2 per cent of the maximum allowed outage length are expected to be taken by RCPs and MFPs respectively in the PWR regardless of other systems. In order to meet a 97 per cent goal capacity factor to be envisaged in a 48 month operating cycle, it is recommended that various possible actions be devised for achieving the higher RCP and MFP operational availability through design, monitoring and maintenance.

scholarly journals Investigation Of Rod Control System Reliability Of Pwr Reactors

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Syaiful Bakhri
Keyword(s):  
Control System ◽  
System Reliability ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Fault Tree ◽  
Reactor Core ◽  
Pressurized Water ◽  
Control Rod ◽  
Control Rods

<p class="NoSpacing1"><span lang="IN">The Rod Control System is </span>employed<span lang="IN"> to adjust the position of the control rods in the reactor core </span>which corresponds with <span lang="IN">the thermal power generated in the core </span>as well as <span lang="IN">the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive </span>employs <span lang="IN">magnetic stepping-type mechanism. This </span>type of <span lang="IN">mechanism consists of a pair of circular coils and latch-style jack with the armature. When the </span>electric <span lang="IN">current </span>is <span lang="IN">supplied to the coils sequentially, the control-rods</span>, which <span lang="IN">are held on the drive shaft</span>, can be driven<span lang="IN"> up</span>ward<span lang="IN"> or down</span>ward<span lang="IN"> in increments. </span>This <span lang="IN">sequential current </span>c<span lang="IN">ontrol</span> drive<span lang="IN"> system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or </span>known also as <span lang="IN">the Rod Control System (RCS). The p</span>urpose of this paper is to investigate the RCS reliability <span lang="IN">of APWR </span>using <span lang="IN">the Fault Tree Analysis (FTA)</span> method<span lang="IN"> since </span>the analysis of reliability which considers<span lang="IN"> the FTA</span> for common CRDM <span lang="IN">can </span>not <span lang="IN">be found</span> in <span lang="IN">any </span>public references. <span lang="IN">The FTA method is used to model the system reliability by developing the fault tree diagram of the system. </span>The<span lang="IN"> results show that the failure of the system is very dependent on the failure of most of the individual systems. However, the failure of the system does not affect the safety of the reactor, since the reactor trips immediately if the system fails. The evaluation results also indicate that the Distribution Panel is the most critical component in the system.</span></p>

Lessons Learned in Implementation of Guidelines for Quantifying Support System Initiating Event Frequencies

2008 ◽  
Author(s):  
Rupert A. Weston ◽  
Ashley J. Mossa
Keyword(s):  
Support System ◽  
Nuclear Power ◽  
Power Plants ◽  
Fault Tree ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Common Cause Failures ◽  
Areas Of Concern ◽  
Pilot Implementation ◽  
American Society

The pilot implementation results for Regulatory Guide 1.200 identified four probabilistic risk assessment (PRA) technical elements that required additional guidance. One of these elements involved the use of fault tree technique to quantify the frequencies of support system initiating events (SSIEs). To address this technical element, guidelines were developed by the Electric Power Research Institute (EPRI) to provide a common industry approach for addressing the identification and quantification of SSIEs. The EPRI guidelines were issued as an interim report to allow trial use and pilot implementation by the industry prior to finalizing the guidelines. These interim guidelines provide an industry-consensus approach for addressing areas of concern in the development of support system initiating event models to ensure that the associated supporting requirements of the American Society of Mechanical Engineers (ASME) PRA Standard for internally initiated events are satisfied. A Pressurized Water Reactor Owners Group (PWROG) pilot implementation of the EPRI interim guidelines was conducted to determine whether the pilot participants have adequately addressed all areas of concern in the development of SSIE models. To determine this, a SSIE model currently used was selected by each of four the pilot participants and subject to detail review to demonstrate whether these models meet the expectations of the EPRI interim guidelines. The EPRI interim guidelines identified the areas of concern to be addressed in using fault tree technique to develop and quantify SSIE models. The guidelines addressed several areas of concern including the treatment of passive failures, the assignment of an appropriate mission time for primary and secondary failures, treatment of common cause failures (CCFs) between running and standby equipment, and consideration of all combinations of CCFs. The PWROG pilot implementation of the interim guidelines summarized the lessons learned and provided feedback to EPRI for consideration in finalizing the guidelines. In addition to the compilation of lessons learned, the PWROG implementation of the EPRI interim guidelines identified existing practices used to develop fault tree models for quantifying SSIE frequencies. Such practices did not necessarily follow a common approach and did not fully meet the expectations of the interim guidelines. Detailed reviews of the SSIE models currently in use at nuclear power plants (NPPs) for the pilot participants demonstrated that the elements of evaluation described in the interim guidelines were not addressed consistently among the PWROG pilot participants. Recommended improvements were identified and incorporated in the SSIE models to meet the expectations of the EPRI interim guidelines. The re-quantification of SSIE frequencies based on the recommended improvements, demonstrated that by not adequately addressing all elements in the evaluation, the SSIE frequency may be under-estimated.

scholarly journals Use of Reliability Block Diagram and Fault Tree Techniques in Reliability Analysis of Emergency Diesel Generators of Nuclear Power Plants

2019 ◽  
Vol 4 (4) ◽  
pp. 814-823 ◽  
Author(s):  
Vanderley de Vasconcelos ◽  
Wellington Antonio Soares ◽  
Antônio Carlos Lopes da Costa ◽  
Amanda Laureano Raso
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Block Diagram ◽  
Electrical Power ◽  
Fault Tree ◽  
Risk Level ◽  
Reliability Block Diagram ◽  
Advantages And Disadvantages ◽  
Diesel Generators

Nuclear power plants (NPPs) are subjected to events such as equipment failures, human errors and common-cause failures, in an environment of complex maintenance, inspection and testing managements. These events will affect the reliability of safety-related systems, as well as the risk level of the plant. Reliability block diagram (RBD) is often used to analyze the effect of item failures on system availability, taking into account their physical arrangement in the system. Fault tree (FT) is a commonly used technique for analyzing risk and reliability in nuclear, aeronautical and chemical industries. It represents graphically the basic events that will cause an undesired top event. Loss of electrical power is one of the main events that influences safe operation of NPPs, as well as accident prevention and mitigation. In case of unavailability of offsite power, emergency diesel generators (EDGs) supply onsite electrical power. This paper carries out reliability analyses of EDGs of NPPs using both RBD and FT techniques. Each technique has its own advantages and disadvantages, allowing a variety of qualitative and quantitative analyses. Outcomes using these two techniques are compared for a typical NPP EDG system.

Detailed Nuclear Power Plant Model for Power System Analysis Based on Matlab-Simulink

2013 ◽  
Vol 291-294 ◽  
pp. 561-570 ◽  
Author(s):  
Jin Tan ◽  
Yue Feng Huang ◽  
Zheng Xu
Keyword(s):  
Power Systems ◽  
Power System ◽  
Nuclear Power ◽  
System Analysis ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Power Generating Unit ◽  
Long Term Stability ◽  
Load Following ◽  
Single Machine Infinite Bus

To research the load-following capability of the nuclear power generating unit, this paper proposed a detailed mathematical model of the pressurized water reactor (PWR) which is suitable for medium- and long-term stability analysis of power systems. Analyzed the interactions between the nuclear power generating unit and the power system, through the simulations of a single machine infinite bus (SMIB) system. The results show that PWR nuclear power generating unit can meet load following requirements to some degree.

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