Application of Seismic Structure Research in Seismic Safety Evaluation of Zhangzhou Nuclear Power Plant

2019 ◽  
Vol 09 (09) ◽  
pp. 790-798
Author(s):  
伟 陈
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Evaluation ◽  
Seismic Structure ◽  
Seismic Safety ◽  
Seismic Safety Evaluation ◽  
Structure Research

Code Developing of Fragility Analysis Program for Seismic Safety Evaluation of Nuclear Power Plant

2017 ◽  
Author(s):  
Zhao Wang ◽  
Jianfeng Yang ◽  
Weijin Wang ◽  
Bingchen Feng ◽  
Xiaoming Zhang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Safety Evaluation ◽  
Seismic Fragility ◽  
Seismic Safety ◽  
Failure Frequency ◽  
Logic Operation ◽  
Seismic Safety Evaluation

For seismic safety evaluation method of nuclear power plant, nuclear power plant seismic margin analysis (SMA) and nuclear power plant seismic probability safety assessment (SPSA) are the most widely used methods. SMA is a method base on deterministic theory. Seismic capacity is valued by high confidence and low failure probability (HCLPF). Through the seismic failure logic of structure, system and components (SSCs), the method can calculated the HCLPF of the whole nuclear power plant, and verify whether the plant can withstand a SSE earthquake test. The SPSA method is the most widely used seismic safety assessment method based on probability theory. Through the analysis and quantification of earthquake accident sequence, a SPSA project can fully identify the seismic risk of nuclear power plant and seismic weak points. Also SPSA can guide the nuclear power plant seismic safety improvement. No matter which method is used to analyze the seismic safety of nuclear power plant, it is necessary to analyze and calculate the seismic fragility of the SSCs. SMA method needs to use a large number of HCLPF data, and seismic fragility analysis and calculation results is one of the main sources of HCLPF data. The SPSA method needs to use seismic fragility data of SSCs which are list in the seismic equipment list (SEL) as input data, so that it can support the quantitative analysis of the risk assessment model. Because of the existence of uncertainty, the seismic fragilities cannot be put to directly logic operation. This brings great difficulty to the using of fragility data. In the paper, the logic operation method and the uncertainty analysis method of seismic fragility is studied, and the calculation program is compiled based on the Monte Carlo method. In this paper, a program is used to calculate the case. The performance of the program is verified and the uncertainty of the system fragility is analyzed. Due to the existence of uncertainty, the fragility cannot put into the numerical calculation directly. In this paper, the calculation method of the failure frequency of components is studied, and the corresponding program is developed by using Monte Carlo method. In this paper, a program is used to calculate the failure frequency of the components under different ground motion levels, and the uncertainty of the failure frequency is also studied.

Current status of design basis earthquake level for nuclear power plant sites in several countries

2021 ◽  
Author(s):  
Hoseon Choi ◽  
Seung Gyu Hyun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
South Korea ◽  
Seismic Activity ◽  
Nuclear Power ◽  
Current Status ◽  
Seismic Safety ◽  
Design Basis ◽  
Design Earthquake ◽  
Design Construction

<p>According to strict criteria step by step for site selection, design, construction and operation, the seismic safety of nuclear power plant (NPP) sites in South Korea are secured by considering design basis earthquake (DBE) level capable of withstanding the maximum ground motions that can occur on the site. Therefore, it is intended to summarize DBE level and its evaluation details for NPP sites in several countries.</p><p>Similar but different terms are used for DBE from country to country, i.e. safe shutdown earthquake (SSE), design earthquake (DE), SL2, Ss, and maximum calculated earthquake (MCE). They may differ when applied to actual seismic design process, and only refer to approximate comparisons. This script used DBE as a representative term, and DBE level was based on horizontal values.</p><p>The DBE level of NPP sites depends on seismic activity of the area. Japan and Western United States, where earthquakes occur more frequently than South Korea, have high DBE values. The DBE level of NPP sites in South Korea has been confirmed to be similar or higher compared to that of Central and Eastern Unites Sates and Europe, which have similar seismic activity.</p>

Critical Section Selection Methodology for the U.S. EPR™ Standard Nuclear Power Plant

2010 ◽  
Author(s):  
Se-Kwon Jung ◽  
Adam Goodman ◽  
Joe Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Critical Section ◽  
Selection Methodology ◽  
Critical Sections ◽  
The U.S

This paper presents a three-tier, critical section selection methodology that is used to identify critical sections for the U.S. EPR™ Standard Nuclear Power Plant (NPP). The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPR™ design; their structural design adequacy provides reasonable assurance of overall U.S. EPR™ structural design adequacy.

Seismic Analysis for an Explosion-Proof Valve Used in Nuclear Power Plant

2017 ◽  
Author(s):  
Juan Luo ◽  
Jiacheng Luo ◽  
Lei Sun
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Seismic Analysis ◽  
Nuclear Safety ◽  
Seismic Load ◽  
Seismic Safety ◽  
Stress Evaluation ◽  
Safety Standards ◽  
Stress And Deformation

Nuclear class equipment should be assessed for seismic safety before they are used in nuclear power plant. According to nuclear safety codes and regulations, all seismic category I equipments shall be designed enduring safety shutdown earthquake (SSE). That is, the stress evaluation needs to be accomplished for those structures. For some components, the deformation evaluation needs to be performed as well to assure the function integrity of the equipment. In this paper, the seismic analysis for an explosion-proof valve used in nuclear power plant, which exactly belongs to seismic category I equipment, has been conducted based on finite element method. The natural frequency, vibration mode and seismic response of the structure have been obtained through calculation, and the stress and deformation under the combined loadings of gravity, internal pressure, blast and seismic load have been evaluated according to ASME AG-1. The bolts of the structure have been qualified according to ASME III-NF as well. The results show that the design of the explosion-proof valve is in compliance with the requirement of corresponding nuclear safety standards.

Probabilistic seismic safety study of an existing nuclear power plant

1980 ◽  
Vol 59 (2) ◽  
pp. 315-338 ◽  
Author(s):  
R.P. Kennedy ◽  
C.A. Cornell ◽  
R.D. Campbell ◽  
S. Kaplan ◽  
H.F. Perla
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Study ◽  
Seismic Safety

Countermeasures Derived From the Lessons of the Fukushima Daiichi Nuclear Power Plant Accident

2013 ◽  
Author(s):  
Tadashi Narabayashi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Pacific Ocean ◽  
Electric Power ◽  
Nuclear Power ◽  
Safety Evaluation ◽  
Lessons Learned ◽  
The Other ◽  
Nuclear Accidents ◽  
Fukushima Daiichi

On March 11, 2011, Tokyo Electric Power Company’s Fukushima Daiichi Nuclear Power Plant (NPP) was hit by a tsunami caused by the Tohoku-Pacific Ocean Earthquake, resulting in nuclear accidents in Units #1 to #4. With the aim of improving the safety of NPPs worldwide, we summarize the lessons that have been learned following a thorough analysis of the event and make specific proposals for improving the safety of such facilities. The author has been involved in investigating the causes of the accidents and developing countermeasures for other NPPs in Japan as a member of the Committee for the Investigation of Nuclear Safety of the Atomic Energy Society of Japan [1], an advisory meeting member of NISA with regard to technical lessons learned from the Fukushima Daiichi NPP accidents, and a Safety Evaluation Member of NISA for the other NPPs in Japan [2].

Approach to Assess Influence of Earthquake-Induced Slope Collapse on Nuclear Power Plant Facilities

2018 ◽  
Vol 12 (04) ◽  
pp. 1841011
Author(s):  
Susumu Nakamura ◽  
Ikumasa Yoshida ◽  
Masuhiro Beppu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Limit State ◽  
Landslide Risk ◽  
Total Probability ◽  
Assessment Procedure ◽  
Limit States ◽  
Seismic Safety ◽  
Slope Collapse

As a result of the disaster of nuclear power plant caused by the 2011 off the pacific coast of Tohoku Earthquake, establishment of a method to estimate the influence of slopes on the seismic safety of nuclear facilities has become necessary. The creation of such a method can yield important information regarding potential risk as well as risk management regarding seismic safety. The existing guidelines used to evaluate landslide risk provide guidance for landslide zoning as well as how landslide risk can be reduced and avoided. According to these guidelines, either people or houses are typically used as targets of risk evaluation. Particularly, for a specific slope, it is necessary to evaluate the damage of the potentially affected structures quantitatively and systematically. Therefore, the definition and basic assessment procedure of three limit states (stability limit, reachable limit and damage limit) are herein described. Furthermore, an evaluation case for a slope model describes the influence of slope collapse due to an earthquake. In this case, the fragility curves, as well as the occurrence probability for each limit state are described and an evaluation example is provided. Regarding new ideas and methods to evaluate the conditional reachable probability and the conditional damage probability as well as a method to evaluate the total probability of all three limit states are proposed. From the results obtained in our example case, it is found that systematical assessment of the risk information of facilities damaged due to slope collapse is useful, and is made possible via numerical analysis.

Critical Section Selection Methodology for the U.S. EPRTM Standard Nuclear Power Plant

2013 ◽  
Vol 135 (4) ◽  
Author(s):  
Se-Kwon Jung ◽  
Adam Goodman ◽  
Joseph Harrold ◽  
Nawar Alchaar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Design ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Safety Evaluation ◽  
Critical Section ◽  
Selection Methodology ◽  
Critical Sections ◽  
The U.S

A three-tier critical section selection methodology that is used to identify critical sections for the U.S. EPRTM standard nuclear power plant (NPP) is presented. The critical section selection methodology includes three complementary approaches: qualitative, quantitative, and supplementary. These three approaches are applied to Seismic Category I structures in a complementary fashion to identify the most critical portions of the building whose structural integrity needs to be maintained for postulated design basis events and conditions. Once the design of critical sections for a particular Seismic Category I structure is complete, the design for that structure is essentially complete for safety evaluation purposes. Critical sections, taken as a whole, are analytically representative of an “essentially complete” U.S. EPRTM design; their structural design adequacy provides reasonable assurance of overall U.S. EPRTM structural design adequacy.

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