scholarly journals BEPU Analysis in LBLOCA Safety Review Calculation

2021 ◽  
Vol 13 (24) ◽  
pp. 14042
Author(s):  
Wei Sun ◽  
Chao Xu ◽  
Yi-Zhen Wang ◽  
Sui-Zheng Qiu ◽  
Yu-Sheng Liu ◽  
...  
Keyword(s):  
Sample Size ◽  
Nuclear Power ◽  
Power Plants ◽  
Correlation Coefficients ◽  
Safety Analysis ◽  
Accurate Estimation ◽  
Loss Of Coolant Accident ◽  
Design Changes ◽  
License Application ◽  
Evaluation Of Parameters

Deterministic safety analysis (DSA) is essential for nuclear power plant licensing. The conservative method followed CFR50 Appendix K, which will lead to a large margin. As one of the DSA methodologies, best estimate plus uncertainty (BEPU) generates more realistic results that can be used in the license application of nuclear power plants (NPPs). However, uncertainty evaluation of parameters is needed in BEPU. In this article, the safety regulatory focuses on the large break loss of coolant accident (LBLOCA) of an advanced PWR. The BEPU analysis is mainly performed by TRACE V5.0 patch 4 code, and the uncertainty analysis is conducted based on DAKOTA code. For correlation coefficients analysis, the sample size is enlarged reasonably. According to the results, this NPP meets the acceptance criteria effectively in LBLOCA with enough margin. By statistic assessment, the set of PCTs calculated has typical normal distribution characters. Based on BEPU, the uncertainties of parameters are studied. Additionally, the influence of sample size on the correlation of parameters is considered too. It could be seen that more samples could permit a more accurate estimation for Spearman partial correlation coefficient (abbreviated as SPCC). The conclusions of this article can provide technical support for the subsequent review of the safety analysis report and the design changes of NPPs.

Sobol′ Sensitivity Analysis Using a Neural Network Model of a LB-LOCA in the ZION Nuclear Power Plant With CATHARE-2 V2.5 Code

2013 ◽  
Author(s):  
Fabrice Fouet ◽  
Pierre Probst
Keyword(s):  
Neural Network ◽  
Sensitivity Analysis ◽  
Nuclear Power ◽  
Uncertainty Propagation ◽  
Correlation Coefficients ◽  
Computer Code ◽  
Numerical Application ◽  
Loss Of Coolant Accident ◽  
Input Parameters ◽  
Artificial Neural Network Ann

In nuclear safety, the Best-Estimate (BE) codes may be used in safety demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters, which comes mainly from the lack of knowledge of the input parameters, is evaluated by estimating the 95% percentile with a high degree of confidence. IRSN, technical support of the French Safety Authority, developed a method of uncertainty propagation. This method has been tested with the BE code used is CATHARE-2 V2.5 in order to evaluate the Peak Cladding Temperature (PCT) of the fuel during a Large Break Loss Of Coolant Accident (LB-LOCA) event, starting from a large number of input parameters. A sensitivity analysis is needed in order to limit the number of input parameters and to quantify the influence of each one on the response variability of the numerical model. Generally, the Global Sensitivity Analysis (GSA) is done with linear correlation coefficients. This paper presents a new approach to perform a more accurate GSA to determine and to classify the main uncertain parameters: the Sobol′ methodology. The GSA requires simulating many sets of parameters to propagate uncertainties correctly, which makes of it a time-consuming approach. Therefore, it is natural to replace the complex computer code by an approximate mathematical model, called response surface or surrogate model. We have tested Artificial Neural Network (ANN) methodology for its construction and the Sobol′ methodology for the GSA. The paper presents a numerical application of the previously described methodology on the ZION reactor, a Westinghouse 4-loop PWR, which has been retained for the BEMUSE international problem [8]. The output is the first maximum PCT of the fuel which depends on 54 input parameters. This application outlined that the methodology could be applied to high-dimensional complex problems.

Sensitivity Analyses for a PWR SCC Case Using the Probabilistic Loss-of-Coolant-Accident (Pro-LOCA) Software

2016 ◽  
Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Design Criterion ◽  
Mitigation Strategies ◽  
Sensitivity Analyses ◽  
Base Case ◽  
Piping System ◽  
Loss Of Coolant Accident ◽  
Piping Systems

As a means of demonstrating compliance with the United States Code of Federal Regulations 10CFR50 Appendix A, General Design Criterion 4 (GDC-4) requirement that primary piping systems for nuclear power plants exhibit an extremely low probability of rupture, probabilistic fracture mechanics (PFM) software has become increasingly popular. One of these PFM codes for nuclear piping is Pro-LOCA which has been under development over the last decade. Currently, Pro-LOCA is being enhanced under an international cooperative program entitled PARTRIDGE-II (Probabilistic Analysis as a Regulatory Tool for Risk-Informed Decision GuidancE - Phase II). This paper focuses on the use of a pre-defined set of base-case inputs along with prescribed variation in some of those inputs to determine a comparative set of sensitivity analyses results. The benchmarking case was a circumferential Primary Water Stress Corrosion Crack (PWSCC) in a typical PWR primary piping system. The effects of normal operating loads, temperature, leak detection, inspection frequency and quality, and mitigation strategies on the rupture probability were studied. The results of this study will be compared to the results of other PFM codes using the same base-case and variations in inputs. This study was conducted using Pro-LOCA version 4.1.9.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

2017 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhang Zhijian ◽  
Tian Zhao-fei ◽  
A. M. Mustafa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Simulation ◽  
Primary System ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping-equipment-structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.

Canadian Nuclear Safety Commission: Readiness to Regulate SMRs in Canada

2011 ◽  
Author(s):  
S. Herstead ◽  
M. de Vos ◽  
S. Cook
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Safety Analysis ◽  
Reactor Power ◽  
Nuclear Safety ◽  
Regulatory Framework ◽  
Nuclear Industry ◽  
Plant Design ◽  
Regulatory Documents

The success of any new build project is reliant upon all stakeholders — applicants, vendors, contractors and regulatory agencies — being ready to do their part. Over the past several years, the Canadian Nuclear Safety Commission (CNSC) has been working to ensure that it has the appropriate regulatory framework and internal processes in place for the timely and efficient licensing of all types of reactor, regardless of size. This effort has resulted in several new regulatory documents and internal processes including pre-project vendor design reviews. The CNSC’s general nuclear safety objective requires that nuclear facilities be designed and operated in a manner that will protect the health, safety and security of persons and the environment from unreasonable risk, and to implement Canada’s international commitments on the peaceful use of nuclear energy. To achieve this objective, the regulatory approach strikes a balance between pure performance-based regulation and prescriptive-based regulation. By utilizing this approach, CNSC seeks to ensure a regulatory environment exists that encourages innovation within the nuclear industry without compromising the high standards necessary for safety. The CNSC is applying a technology neutral approach as part of its continuing work to update its regulatory framework and achieve clarity of its requirements. A reactor power threshold of approximately 200 MW(th) has been chosen to distinguish between large and small reactors. It is recognized that some Small Modular Reactors (SMRs) will be larger than 200 MW(th), so a graded approach to achieving safety is still possible even though Nuclear Power Plant design and safety requirements will apply. Design requirements for large reactors are established through two main regulatory documents. These are RD-337 Design for New Nuclear Power Plants, and RD-310 Safety Analysis for Nuclear Power Plants. For reactors below 200 MW(th), the CNSC allows additional flexibility in the use of a graded approach to achieving safety in two new regulatory documents: RD-367 Design of Small Reactors and RD-308 Deterministic Safety Analysis for Small Reactors. The CNSC offers a pre-licensing vendor design review as an optional service for reactor facility designs. This review process is intended to provide early identification and resolution of potential regulatory or technical issues in the design process, particularly those that could result in significant changes to the design or analysis. The process aims to increase regulatory certainty and ultimately contribute to public safety. This paper outlines the CNSC’s expectations for applicant and vendor readiness and discusses the process for pre-licensing reviews which allows vendors and applicants to understand their readiness for licensing.

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