Safety Analysis on Mitigation Overpressure ATWS in VVER-1000 NPP

2013 ◽  
Author(s):  
Ting Qi ◽  
Changjiang Yang
Keyword(s):  
Nuclear Power ◽  
Late Phase ◽  
Safety Analysis ◽  
Severe Accident ◽  
Injection System ◽  
Feed Water ◽  
System Pressure ◽  
Design Characteristics ◽  
Coolant System ◽  
Relap5 Code

Safety analysis on operating nuclear power plant (NPP) plays an important role on nuclear energy application, especially after the severe accident of Fukushima plants. This paper focuses on one of the operating NPPs in China, the TianWan NPP Unit1&2. TianWan Unit1&2 belong to VVER-1000 NPP type, which have special design characteristics and different safety migrating methods comparing with other domestic PWR NPPs in China. Calculations and analyses were made to give thermal hydraulics support to Level-1 PSA of this VVER type PWR on the anticipated transient without scram (ATWS) accidents. The calculation of loss of main feed water ATWS was carried out using the RELAP5 code to evaluate the intrinsic safety mainly impacted by fuel and moderator. The model also considers reactivity introduction caused by the change of boron concentration to find out the influence of the emergency boron injection system (JDH) on mitigating the coolant system over pressure. The paper gives out the success criteria of the safety valves of pressurizer (PRZ) with the critical moderator temperature coefficient (MTC) value according to the condition of coolant system pressure being under the pressure limit. It is indicated that the JDH system can play an important role on mitigating the over pressure of coolant system in the late phase of the transient.

Development and Verification of Severe Accident Simulation Capability for a NPP Simulator

2013 ◽  
Author(s):  
Zhifei Yang ◽  
Xiaofei Xie ◽  
Xing Chen ◽  
Shishun Zhang ◽  
Yehong Liao ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation System ◽  
Severe Accident ◽  
Normal Operation ◽  
Feed Water ◽  
Accident Simulation ◽  
Accident Conditions ◽  
Relap5 Code

It is reflected in the severe accident in Fukushima Daiichi that the emergency capacity of nuclear power plant needs to be enhanced. A nuclear plant simulator that can model the severe accident is the most effective means to promote this capacity. Until now, there is not a simulator which can model the severe accident in China. In order to enhance the emergency capacity in China, we focus on developing a full scope simulator that can model the severe accident and verify it in this study. The development of severe accident simulation system mainly includes three steps. Firstly, the integral severe accident code MELCOR is transplanted to the simulation platform. Secondly, the interface program must be developed to switch calculating code from RELAP5 code to MELCOR code automatically when meeting the severe accident conditions because the RELAP5 code can only simulate the nuclear power plant normal operation state and design basis accident but the severe accident. So RELAP5 code will be stopped when severe accident conditions happen and the current nuclear power plant state parameters of it should be transported to MELCOR code, and MELCOR code will run. Finally, the CPR1000 nuclear power plant MELCOR model is developed to analyze the nuclear power plant behavior in severe accident. After the three steps, the severe accident simulation system is tested by a scenario that is initiated by the station black out with reactor cooling pump seal leakage, HHSI, LHSI and auxiliary feed water system do not work. The simulation result is verified by qualitative analysis and comparison with the results in severe accident analysis report of the same NPP. More severe accident scenarios initiated by LBLOCA, MBLOCA, SBLOCA, SBO, ATWS, SGTR, MSLB will be tested in the future. The results show that the severe accident simulation system can model the severe accident correctly; it meets the demand of emergency capacity promotion.

The Severe Accident Analysis of CPR1000 Based on MELCOR Code

2013 ◽  
Author(s):  
Longze Li ◽  
Mingjun Wang ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Failure Time ◽  
Water Pressure ◽  
Severe Accident ◽  
Feed Water ◽  
Calculation Results ◽  
Station Blackout ◽  
Accident Response ◽  
Basic Standard

The severe accident of CPR1000 caused by station blackout with the SG safety valve failure is simulated and analyzed using MELCOR code in this work. The CPR1000 power plant severe accident response process and the results with three different assumptions, which are no the seal leakage nor the auxiliary feed water, the seal leakage and auxiliary feed water exist, the seal leakage exist but no auxiliary feed water separately, are analyzed. According to the calculation results, without the seal leakage and auxiliary feed water, pressure vessel would fail at 9576 s. When auxiliary feed water was supplied, pressure vessel’s failure time would delay nearly 30000s. When the seal leakage exists, pressure vessel’s failure time would delay about 50 s. The results are meaningful and significant for comprehending the detailed process of severe accident for CPR1000 nuclear power plant, which is the basic standard for establishing the severe accident management guideline.

Analysis of Early Stage Accident at TEPCO’s Fukushima Daiichi NPS

2013 ◽  
Author(s):  
Hiroshi Ono ◽  
Hideo Konishi
Keyword(s):  
Nuclear Power ◽  
Early Stage ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
System Behavior ◽  
Power Station ◽  
Detailed Simulation ◽  
Fukushima Daiichi ◽  
Relap5 Code ◽  
Plant Control

The operation of the Isolation Condenser (IC) played an important role in the progression of the accident at the unit 1 of Fukushima Daiichi nuclear power station. Analyses of Unit 1 accident in the early stage (prior to the occurrence of core melt) were performed using plant dynamics analysis code RELAP5/MOD31 and the results were compared with measured data. In the RELAP5 code, analysis scope of target is not a severe accident. But, the detailed simulation in the early stage of accident including the plant control system behavior is possible. Moreover, some sensitivity analyses were conducted and the reactor behavior under IC operations was examined.

Level 2 PSA Overview of HPR1000 Nuclear Power Plant

2020 ◽  
Author(s):  
Wang Ziguan ◽  
Lu Fang ◽  
Yang Benlin ◽  
Chen Shi ◽  
Hu Lingsheng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Injection System ◽  
Risk Level ◽  
Level 2

Abstract Risk-informed design approaches are comprehensively implemented in the design and verification process of HPR1000 nuclear power plant. Particularly, Level 2 PSA is applied in the optimization of severe accident prevention and mitigation measures to avoid the extravagant redundancy of system configurations. HPR1000 preliminary level 2 PSA practices consider internal events of the reactor in the context of at-power condition. Severe accidents mitigation and prevention system and its impact on the overall large release frequency (LRF) level are evaluated. The results showed that severe accident prevention and mitigation systems, such as fast depressurization system, the cavity injection system and the passive containment heat removal system perform well in reducing LRF and overall risk level of HPR1000 NPP. Bypass events, reactor rapture events, and the containment bottom melt-through induced by MCCI are among the dominant factors of the LRF. The level 2 PSA analysis results indicate that HPR1000 design is reliable with no major weaknesses.

Transient Analysis of Dedicated Pressure Relief Valve Under High-Pressure Condition in Severe Accident for Tianwan 56 Nuclear Power Plant

2020 ◽  
Author(s):  
Youyou Xu ◽  
Jian Deng ◽  
Xiaoji Wang ◽  
Lingjun Wu ◽  
Ming Zhang ◽  
...  
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Feed Water ◽  
Pressure Condition ◽  
Relief Valve ◽  
Pressure Relief ◽  
Accident Sequences ◽  
Water Accident

Abstract In the management of severe accident of nuclear reactor, the pressure relief of reactor coolant system (RCS) is an important mitigation measure to prevent high pressure core melt (HPCM). In the safety system improvement of Tianwan56 nuclear power plant, the optimization measure of adding the dedicated pressure relief valve (DPRV) for severe accident were adopted. This improvement allows the reactor to release the pressure of RCS before the reactor vessel being damaged to mitigate the consequence of reactor melt accident under high-pressure condition. Based on the analysis of severe accident sequences, the total loss of feed water accident is confirmed to cover the various severe accident consequences which may lead to HPCM accident. This paper studied the transient characteristics of total loss of feed water accident sequences, and the factors such as valve opening delay on the operating temperature of the valve were researched. Finally, the representative and envelope operating condition of DPRV under severe accident was clarified. Besides, the temperature curve of fluid passing through the valve and the maximum temperature the valve experienced were obtained. This research provides the valuable and indispensable basis to the operability and integrity analysis of DPRV in severe accident.

Application of Level 2 PSA in the Design of Cavity Injection System for Nuclear Power Plant

2018 ◽  
Author(s):  
Wentao Zhu ◽  
Wenjing Li
Keyword(s):  
Decision Making ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Assessment Tool ◽  
Quantitative Risk Assessment ◽  
Severe Accident ◽  
Injection System ◽  
Retention Strategy ◽  
Level 2

After Fukushima nuclear power plant accident, severe accident is getting more and more concerns all over the world. In order to mitigate severe accident and improve the safety of nuclear power plant, two different strategies are applied in different plants. One is in-vessel melt retention strategy, and the other is ex-vessel melt retention strategy. Tianwan nuclear power plant is an improved Gen II nuclear power plant and in-vessel melt retention strategy is adopted in the plant. In order to achieve this strategy, cavity injection system is designed for the plant. Probabilistic Safety Analysis is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. For this plant, in order to optimize the design of cavity injection system, improve the safety level of nuclear power plant, and meanwhile, improve the engineering implementation and economization, Level 2 PSA was used for this decision-making process. In this paper, the Level 2 PSA for this plant and the application for the design of cavity injection system are introduced.

scholarly journals Evaluation of the Potential for Containment Bypass due to Steam Generator Tube Rupture in VVER-1000/V320 Reactor during Extended SBO sequence using SCDAP/RELAP5 code

2021 ◽  
Vol 9 (4) ◽  
pp. 9-15
Author(s):  
Van Thai Nguyen ◽  
Manh Long Doan ◽  
Chi Thanh Tran
Keyword(s):  
Steam Generator ◽  
Natural Circulation ◽  
Severe Accident ◽  
Pressure Relief ◽  
Plant Behavior ◽  
Pressure Relief Valves ◽  
Station Blackout ◽  
Coolant System ◽  
Relap5 Code ◽  
Tube Failure

A severe accident-induced of a Steam Generator (SG) tube releases radioactivity from the Reactor Coolant System (RCS) into the SG secondary coolant system from where it may escape to the environment through the pressure relief valves and an environmental release in this manner is called “Containment Bypass”. This study aims to evaluate the potential for “Containment Bypass” in VVER/V320 reactor during extended Station Blackout (SBO) scenarios that challenge the tubes by primarily involving a natural circulation of superheated steam inside the piping loop and then induce creep rupture tube failure. Assessments are made of SCDAP/RELAP5 code capabilities for predicting the plant behavior during an SBO event and estimates are made of the uncertainties associated with the SCDAP/RELAP5 predictions for key fluid and components condition and for the SG tube failure margins. 

Implicit Fatigue Design of Piping Systems for License Renewal of B31.1 Plants

2003 ◽  
Author(s):  
K. C. Chang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fatigue Analysis ◽  
Fatigue Design ◽  
System Pressure ◽  
Sufficient Detail ◽  
Piping Systems ◽  
Thermal Bending ◽  
License Renewal ◽  
Coolant System

The fatigue analysis of older vintage nuclear power plants was normally performed following the implicit method defined in the USAS B31.1, Power Piping Code. The USAS B31.1 Code applies reduction factors to allowable stresses for loadings with specific numbers of cycles. This effectively reduces stress amplitudes due to thermal bending and prevents fatigue damage from occurring. However, it does not provide sufficient detail to address the environmental impacts on fatigue when license renewal application is prepared for a specific plant. Explicit fatigue analyses were performed for many components in plants whose piping was designed by the implicit fatigue analysis methods. Over the years, modified or improved plant operating procedures may have been implemented. More sophisticated calculations became available for use on several components for which some explicit fatigue analyses is required to address the environmental effects for 60 years of operation. This paper addresses the type of information required to evaluate the effect of environmental factors on fatigue in a license renewal application for plants with reactor coolant system pressure boundary piping designed to B31.1 Code.

Study on the Application of CDAG Methodology in EPR Nuclear Power Plant

2017 ◽  
Author(s):  
Yin Yuhao ◽  
Huang Yichao ◽  
Zhao Feng
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fission Product ◽  
Damage Assessment ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Damage ◽  
Calculation Results ◽  
Design Characteristics ◽  
European Nuclear

The Westinghouse Owners Group Core Damage Assessment Guidance (CDAG), which has been authorized by the NRC staffs, is now used by licensee emergency response organization staff for estimating the extent of core damage that may have occurred during an accident at a Westinghouse nuclear power plant. On the other hand, EPR is a 3rd generation nuclear power plant, which applies the advanced European nuclear power technology. This paper introduced Core Damage Assessment Guidance methodology in detail. The CDAG methodology is then attempted to apply to the EPR nuclear power plant. Detailed calculations have been performed for the setpoints of containment radiation monitors (CRM) and core exit thermocouples (CETs) with EPR design characteristics, which are the two main methods for estimation core damage amount. This paper also focuses the discussion on the reasons of difference of core damage estimating results between CRM method and CETs method; based on the discussion, several advices are provided when the two methods show a reasonable discrepancy in conclusions. Several conclusions can be made from the discussions in this article that 1)the Westinghouse Owners Group CDAG methodology proved to be reasonable when applied to EPR power plant for core damage assessment under severe accident; 2) the CDAG methodology which reflect the latest understanding of fission product behavior, is very simple and timely for core damage assessment based on NPP (nuclear power plant) real-time parameters; 3) conservative calculation results of setpoints on CRM and CETs based on EPR design show a reasonable trend and range; 4) it is concluded that several factors such as the releasing way, RCS fission product retention, fuel burnups might have great impact on the estimating results, when the results from two main indications (CRM and CETs) show an unexpected response.

Analysis of Severe Accidents in VVER-1000 Reactors Using the Integral Code ASTEC

2009 ◽  
Author(s):  
Polina Tusheva ◽  
Nils Reinke ◽  
Eberhard Altstadt ◽  
Frank Schaefer ◽  
Frank-Peter Weiss ◽  
...  
Keyword(s):  
Power System ◽  
Failure Time ◽  
Late Phase ◽  
Electric Power System ◽  
Physical Models ◽  
Severe Accident ◽  
Feed Water ◽  
The Core ◽  
Lower Head ◽  
Station Blackout

The studies presented are aiming at a detailed investigation of the behaviour of a VVER-1000/V-320 reactor and the containment structures during a postulated severe accident, including the ways and means by which these accidents may be prevented or mitigated. A hypothetical station blackout scenario (loss of the offsite electric power system concurrent with a turbine trip and unavailability of the emergency AC power system), belonging to the typical beyond design basis accidents, has been investigated. Station blackout results in reactor shut down, loss of feed water and trip of all reactor coolant pumps. Continuous evaporation of the secondary side leads to steam generators’ depletion followed by heating up of the core. In case of unavailability of essential safety systems the core will be severely damaged and finally the reactor pressure vessel (RPV) might fail. The analyses are performed using the integral code ASTEC commonly developed by IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and GRS (Gesellschaft fu¨r Anlagen- und Reaktorsicherheit mbH). Code-to-code comparative analyses for the early thermal-hydraulic phase have been performed with the GRS code ATHLET. A large number of sensitivity calculations have been done regarding the axial core power distribution, heat losses, and RPV lower head modelling. The analyses have shown that, despite the considerable differences in the codes themselves, the calculation results are similar in terms of thermal hydraulic response. There are discrepancies in timings of phenomena, which are within the limitations of the physical models and the applied nodalizations. It was one objective of this investigation to evaluate the Severe Accident Management (SAM) procedures for VVER-1000 reactors, by for instance estimating the time available for taking appropriate decisions and preparing counter-measures. To evaluate the effect of possible operator actions, a SAM procedure (primary side depressurization) is included into the simulation. Without SAMs, the simulation provides plastic rupture of the RPV after approximately 4.3 h, while with SAMs, a prolongation of the vessel failure time is obtained by approximately 90 minutes. Currently, the late phase of the accident is investigated in more detail by comparing the lower head behaviour as simulated by ASTEC with results from dedicated finite element calculations. The work contributes to the reliability of the ASTEC code by means of plant applications.

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