Evaluation of Representative Piping Systems Designed by Implicit Fatigue Concept

2008 ◽  
Author(s):  
Shin-Beom Choi ◽  
Sun-Hye Kim ◽  
Yoon-Suk Chang ◽  
Jae-Boong Choi ◽  
Young-Jin Kim ◽  
...  
Keyword(s):  
Fatigue Life ◽  
Nuclear Power ◽  
Heat Removal ◽  
Environmental Water ◽  
Lessons Learned ◽  
Metal Fatigue ◽  
Aging Effects ◽  
Age Related ◽  
Piping Systems ◽  
Surge Line

NUREG-1801 provides generic aging lessons learned to manage aging effects that may occur during continued operation beyond the design life of nuclear power plant. According to this report, the metal fatigue, among several age-related degradation mechanisms, is identified as one of time-limited aging analysis item. The objective of this paper is to introduce fatigue life evaluation of representative surge line and residual heat removal system piping which was designed by implicit fatigue concept. For the back-fitting evaluation employing explicit fatigue concept, detailed parametric CFD as well as FE analyses results are used. The well-known ASME Section III NB-3600 procedure is adopted for the metal fatigue and NUREG/CR-5704 procedure is further investigated to deal with additional environmental water effects. With regard to the environmental effect evaluation, two types of fatigue life correction factors are considered, such as maximum Fen and individual Fen. As a result, it was proven that a thermal stratification phenomenon is the governing factor in metal fatigue life of the surge line and strain rate is the most important parameter affecting the environmental fatigue life of both piping. The evaluation results will be used as technical bases for continued operation of OPR 1000 plant.

Lessons Learned for Nuclear Piping Integrity in New Reactors

2011 ◽  
Author(s):  
G. Wilkowski ◽  
F. Brust ◽  
P. Krishnaswamy ◽  
K. Wichman ◽  
D.-J. Shim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Lessons Learned ◽  
Dynamic Strain ◽  
Piping System ◽  
Aging Effects ◽  
Piping Systems ◽  
Loading Effects

From the early 1980’s to the present time, there has been a significant amount of research and development on the structural integrity of nuclear power plant piping. From those efforts, there are a number of lessons that could be applied to design and fabrication of new nuclear power plant piping systems. In this paper, the various aspects evaluated in NRC-funded efforts for understanding degraded piping were reviewed and implications on how to avoid detrimental aspects were discussed, as well as some more recent efforts. Some of these aspects include; (1) materials aspects (variability of wrought stainless steel base metal toughness with composition, dynamic strain aging effects on toughness of ferritic steels, fracture toughness in HAZ/fusion lines, material anisotropy effects on toughness, effects of static versus dynamic loading on material toughness, cyclic loading effects during seismic loading on toughness, thermal aging effects on strength and toughness), (2) designing weld sequencing to avoid SCC cracking; (3) crack morphology effects on leak-rate evaluations, (4) system effects that can significantly affect the structural integrity analysis of the piping system (secondary stresses, restraint of pressure induced bending, system displacement and rotation constraints, and margins associated from full dynamic analyses).

A New Method to Weaken Thermal Stratification Phenomena in Pipes of Nuclear Power Plant

2013 ◽  
Author(s):  
Shengfei Wang ◽  
Yuxin Pang ◽  
Xiaojing Li ◽  
Dandan Fu ◽  
Yang Li ◽  
...  
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Thermal Stresses ◽  
Heat Pipes ◽  
New Method ◽  
Simple Experiment ◽  
Pressurized Water ◽  
Piping Systems ◽  
Surge Line ◽  
Water Reactors

Thermal stratification phenomena are observed in piping systems of pressurized water reactors, especially in the pressurizer surge line. As a result of the thermal stratification induced thermal stresses, fatigue problems can occur in the pipework. US NRC requirements have also identified flow stratification in surge lines as a phenomenon that must be considered in the design basis of surge lines. In this paper, a new method to reduce thermal stratification is proposed. As we all know, heat pipe is a simple device with no moving parts and can transfer large quantities of heat over fairly large distance. The new method is that using heat pipes to weaken the thermal stratification. In order to validate the new method, a simple experiment and theoretical analysis was taken. The results show that, the temperature difference of thermal stratification with heat pipes is smaller than the stratification without heat pipes. A design scheme was also given at the end of paper.

Assurance of Seismic Safety by Defense-in-Depth and Safety Goals: Lessons Learned From the Chuˆetsu-Oki Earthquake (2007)

2008 ◽  
Author(s):  
G. Saji
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Lessons Learned ◽  
Seismic Events ◽  
Seismic Safety ◽  
Event Scale ◽  
Guiding Principle ◽  
Safety Risks ◽  
Defense In Depth

Although the basic safety concerning control of reactivity, residual heat removal and confinement, was assured in the Kashiwazaki–Kariwa Nuclear Power Plants at the time of the Chuˆetsu-Oki Earthquake (2007), the accident caused great public concern as to the seismic safety of NPPs. The earthquake resulted in severe economic impacts, far exceeding the actually negligible environmental effect. The public is calling for a reassessment of the seismic safety of NPPs, as they are unable to understand the basic safety approaches of the Japanese seismic design practice. The earthquake significantly exceeded the design basis ground motion in all units. Due to this the seismic consequences, especially those malfunctions and damages in the lower seismic class items, are not surprising. The following three topics are highlighted from the lessons learned and the author’s reappraisal of the current seismic safety approach, namely: (1) Prevention of seismic consequences of the lower seismic class items, (2) Measures to ensure the seismic safety by including defense-in-depth, and (3) Reduction of seismic safety risks as low as reasonably achievable referring to ‘safety goals for seismic events.’ The author believes that ‘reduction of seismic risks as low as practical (S-ALAP)’ should be a new guiding principle for lower safety class items, in line with the new concept (gensai) developed in light of the Great Hanshin-Awaji Earthquake (1995), acknowledging difficulties of coping with earthquakes just by conservative design. For a reasonable reduction of seismic risks, it is necessary to answer to question of ‘how safe is safe enough.’ The author developed a safety goal for seismic consequences by integrating the International Nuclear Event Scale (INES) and Farmer’s probabilistic siting criteria. It is an extension of the author’s quantitative safety goals for non-seismic events already published in a series of previous papers (including RESS Vol. 80-2, pp. 143–161, 163–172; PSA’05-139985, 139989, 139990: ICONE14-89351).

CFD Analysis of Thermal Stratification Phenomena in the RNS Suction Line of a Gen-III Nuclear Power Plant

2013 ◽  
Author(s):  
Yan Li ◽  
Daogang Lu ◽  
Zhigang Wang ◽  
Jian Wu ◽  
Fengyun Yu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Heat Removal ◽  
Radial Temperature ◽  
Piping Systems ◽  
Suction Line ◽  
Power Operation

Thermal stratification phenomena in piping systems of nuclear power plant would threaten the structural integrity of pipes, which are caused by the significant change of water density with temperature. To provide temperature gradients for the stress analysis of Normal heat Removal System (RNS) suction line of a Gen-III nuclear power plant, the relevant thermal stratification phenomena are analyzed by CFD in this paper. Cases without leakage (normal power operation) and with leakage are both studied. The results show that the first portion of pipe (one meter or so) near the hot leg is isothermal for normal power operation due to the penetrating flow. In the remaining portion, the radial temperature drops are of the order of 20∼27 K for no leakage case. For the leakage case, the radial temperature drops are 23 K or less, which are relatively smaller than those for the no leakage case due to the net hot flow from the hot leg to the valve.

Use of Markov Piping Reliability Models to Evaluate Time Dependent Frequencies of Loss of Coolant Accidents

2004 ◽  
Author(s):  
Karl N. Fleming ◽  
Bengt O. Y. Lydell
Keyword(s):  
Markov Model ◽  
Nuclear Power ◽  
Interesting Property ◽  
Time Dependent ◽  
Piping System ◽  
Accelerated Corrosion ◽  
Reliability Models ◽  
Age Related ◽  
Loss Of Coolant ◽  
Piping Systems

Markov model theory has been applied to develop a method to evaluate the influence of alternate strategies for in-service inspection and leak detection on the frequency of leaks and ruptures in nuclear power plant piping systems [1–4]. This approach to quantification of pipe rupture frequency was originally based on a Bayes’ uncertainty analysis approach to derive piping system failure rates from a combination of service experience data and some simple reliability models [5–7]. More recently the Markov model approach has been used in conjunction with probabilistic fracture mechanics methods in the study of flow accelerated corrosion [8]. One interesting property of the Markov model is its capability to evaluate time dependent rupture frequencies via the model hazard rate. In this paper this time dependent modeling capability is used to investigate the age related and time dependent frequencies of loss of coolant accident (LOCA) initiating event frequencies. A case is presented that plant age dependent LOCA frequencies should be used in lieu of other metrics commonly used in probabilistic risk assessments and in risk informed inservice inspection evaluations. Such more commonly used metrics include the assumed constant failure rate method and the lifetime average rupture probability. Both of these methods are shown to provide optimistic estimates of LOCA frequencies for plants in the latter part of their design lifetimes, which most operating plants are approaching.

Research on Emergency System of Injection Estimation of Reactor Core Recovery and Reactor Core Decay Heat Removal

2017 ◽  
Author(s):  
Liao Feiye ◽  
Jiang Pingting ◽  
Liu Wang ◽  
He Dongyu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Reactor Core ◽  
Heat Removal ◽  
Lessons Learned ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Core Recovery ◽  
Emergency System

One of the lessons learned from Fukushima accident is that the existing procedures used in Nuclear Power Plants (NPPs) are not executed effectively and quickly enough after such an extended accident, for the accident is complex and people are too nervous in such a situation. Thus, emergency system that helps to raise diagnosis efficiency is necessary. In the paper, a quick diagnosis system on injection estimation of reactor core recovery and decay heat removal injection estimation is developed to meet the urgent needs and strengthen requirements for the training and application among utilities and nuclear regulators. The system will assist regulators to quickly know whether the currently flow will probably recover the reactor core, or whether the current injection capacity is sufficient to quench and recover the reactor core, directly after input present parameters into the system. In the system, Matlab method is used, and intuitive insights are considered, which is propitious to give immediate graphical interface and reduce possibility of human error.

Development of Guideline on Seismic Fragility Evaluation for Aged Piping

2021 ◽  
Author(s):  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Koichi Masaki ◽  
Yinsheng Li
Keyword(s):  
Risk Assessment ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Seismic Fragility ◽  
Degradation Mechanisms ◽  
Probabilistic Fracture Mechanics ◽  
Age Related ◽  
Piping Systems

Abstract The seismic probabilistic risk assessment is an important methodology to evaluate the seismic safety of nuclear power plants. In this assessment, the core damage frequency is evaluated from the seismic hazard, seismic fragilities, and accident sequence. Regarding the seismic fragility evaluation, the probabilistic fracture mechanics can be applied as a useful evaluation technique for aged piping systems with crack or wall thinning due to the age-related degradation mechanisms. In this study, to advance seismic probabilistic risk assessment methodology of nuclear power plants that have been in operation for a long time, a guideline on the seismic fragility evaluation of the typical aged piping systems of nuclear power plants has been developed considering the age-related degradation mechanisms. This paper provides an outline of the guideline and several examples of seismic fragility evaluation based on the guideline and utilizing the probabilistic fracture mechanics analysis code.

A Validation of RELAP on Predicting Nuclear Power Plant Phenomena

2018 ◽  
Author(s):  
Ji Soo Ahn ◽  
Michael Bluck ◽  
Matthew Eaton ◽  
Chris Jackson
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Feed Water ◽  
Multiple Channels ◽  
Horizontal Pipe ◽  
One Dimensional ◽  
Piping Systems ◽  
Surge Line ◽  
Input Model

In this study, RELAP5’s capability to simulate thermal stratification under different conditions is assessed. In nuclear power plants (NPPs), thermal stratification can occur in the following locations: pressurizer, piping systems such as hot legs, cold legs, surge lines, and cooling tanks if available. In general, thermal stratification in a horizontal pipe could not be simulated by RELAP5 due to the inherent one-dimensional setting. Moreover, RELAP5 failed to simulate turbulent penetration which was often a pre-requisite prior to thermal stratification in a pipe. This type of situation could arise in connection between hot leg and surge line, spray lines, feed water lines, etc. It is recommended that for this type of problem CFD be used. In the literature, it was found that RELAP5 was capable of simulating thermal stratification in a pool or a tank-like component if multiple channels and crossflow junctions were used. However, due to uncertainties associated with the input model, the current RELAP5 model failed to reproduce experimental data and therefore further investigation would be required to identify the sources of error.

Time Limited Aging Analyses Requirements for License Renewal for U.S. Nuclear Power Plants

2004 ◽  
Author(s):  
Mansoor H. Sanwarwalla
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Management Program ◽  
Metal Fatigue ◽  
Aging Effects ◽  
Electric Equipment ◽  
Design Calculations ◽  
License Renewal

One of the requirements for license renewal for US nuclear plants stated in the United States Nuclear Regulatory Commission (USNRC) regulations in the License Renewal Rule (LRR) 10CFR Part 54 (Ref. 1) is the identification and updating of Time-Limited Aging Analyses (TLAA). During the design phase for a plant, certain assumptions about the length of time the plant would be operated were made and incorporated into design calculations for several of the plant’s systems, structures and components (SSCs). Examples of TLAAs are analyses of metal fatigue, environmental qualification (EQ) of electric equipment, etc. For a renewed license, these calculations have to be reviewed to verify that they remain valid for the period of extended operation. However, the LRR does allow TLAA-associated aging effects to be managed by an aging management program. This paper discusses the USNRC regulatory requirements for TLAAs and the industry’s response for addressing the TLAAs. It also discusses the issues regarding the generic set of TLAAs that have been identified by the NRC in NUREG-1801 (Ref. 2), and how these have been addressed by all the plants that have received their renewed license. The paper also identifies certain plant specific TLAAs.

Cyclic Loading Test and Numerical Analysis of Flange Joint and Reducer of RHR Piping Systems

2014 ◽  
Author(s):  
Zih-Yu Lai ◽  
Yan-Fang Liu ◽  
Ching-Ching Yu ◽  
Juin-Fu Chai ◽  
Fan-Ru Lin ◽  
...  
Keyword(s):  
Finite Element ◽  
Numerical Analysis ◽  
Cyclic Loading ◽  
Nuclear Power ◽  
Numerical Models ◽  
Heat Removal ◽  
Piping System ◽  
Loading Test ◽  
Element Analysis ◽  
Piping Systems

According to the seismic risk assessment results presented in the Final Safety Analysis Report (FSAR) for a nuclear power plant in Taiwan, the failure of Residual Heat Removal (RHR) piping system occurs in both of the two accident sequences with the highest contributions for core damage. The seismic performance of RHR piping system depends on the capacity of its components, such as supports, flanged joints and reducers. For the need of seismic response-history analysis of RHR piping systems, we developed detailed numerical models of flanged joint and reducer using finite element analysis software (ABAQUS and SAP2000). The proposed finite element models were verified by the experimental results. The pure bending tests with four-point cyclic loading were conducted for sample flanged joint and reducer to investigate their mechanical behaviors. The displacement and rotation responses identified from the tests are in good agreement with the results of numerical analysis. A preliminary simplified model of flanged joints was also proposed in this study to improve the efficiency of numerical analysis for RHR piping system.

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