A Design Proposal and Assessment of In-Vessel Retention Strategy for 2nd Generation PWR

2013 ◽  
Author(s):  
Yabing Li ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Metal Layer ◽  
Success Probability ◽  
Computer Code ◽  
Injection System ◽  
Accident Analysis ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Water Storage Tank ◽  
2Nd Generation ◽  
Probability Density Distributions

In order to satisfy the higher safety requirements after the Fukushima Accident, in-vessel retention (IVR) strategy is suggested to be a measure for the improvement for the operating 2nd generation Pressurized Water Reactor (PWR). In this paper, a design of passive cavity flooding system isproposed for the 2nd generation modified PWR with reference to the design of in-containment refueling water storage tank (IRWST)for AP1000, and the IVR assessmentis evaluated with the risk oriented accident analysis methodology (ROAAM). Seven representational accident sequences are selected for the IVR assessment and analyzed by lumped-parameter safety analysis computer code. Accident analysis focuses on four key parameters forthe assessment, that is, decay power, zirconium oxidation fraction, and both the mass of oxidic pool and metal layer. The values of them are obtained through analysis and used for the assessment. The probability density distributions of those parameters are determined by combining the analysis results and engineering judgment. The success probability of IVR from the viewpoint of thermal failure is predicted using a program written by the Matlab code. Furthermore, some sensitivity analysis and parametric studies are investigated to support the assessment. The assessment result shows that the success probability of the cavity injection system is higher than 99%. Detail analysis about system reliability and feasibility is needed for further work.

Study on Typical Initial Event of ATWS for Small Modular Reactor

2016 ◽  
Author(s):  
Zhang Dan ◽  
Ran Xu ◽  
Qiu Zhifang ◽  
Zhou Ke ◽  
Feng Li
Keyword(s):  
Optimum Design ◽  
Safety Analysis ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Reactor Coolant ◽  
Initial Event ◽  
Small Modular Reactor ◽  
Activation System

The method for ATWS (anticipated transient without scram) analysis was completely developed for commercial pressurized water (PWR) reactor plants, especially for selecting of typical initial events. For accident analysis of ATWS, it is different between PWR and small modular reactor (SMR), as different structures and characters, and it is necessary to study the typical initial events for these reactors. Based on the standard of PWR, the demanding for ATWS analysis was studied and the consequences for typical anticipated transient was calculated using RELAP5/MOD3.2 code, “maintain reactor coolant pressure boundary integrity” was selected as limiting criterion. The results shows for SMR, anticipated transient with the most serious consequence for ATWS are loss of offsite power and inadvertent control rod withdraw event, this conclusion will support to prepare the safety analysis report and optimum design of diversity activation system (DAS) for SMR.

The Analytical and Experimental Research of In-Vessel Corium Retention in CPR1000

2010 ◽  
Author(s):  
Xing Chen ◽  
Shishun Zhang ◽  
Jiming Lin ◽  
Huiyong Zhang
Keyword(s):  
Experimental Research ◽  
Success Probability ◽  
Outer Wall ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Structural Failure ◽  
Pressurized Water ◽  
Thermal Failure ◽  
Near Future ◽  
Preliminary Phase

The analytical and experiment research of In-Vessel Corium Retention (IVR) in the Chinese Pressurized-water Reactor 1000 MWe (CPR1000) are introduced. The IVR research consists of preliminary phase and detailed phase. The analysis of thermal failure, structural failure and penetration failure of Reactor Pressure Vessel (RPV) and the experimental research of External Reactor Vessel Cooling (ERVC) are performed at preliminary phase. Analysis results show that the RPV failure is the dominated by thermal failure mode and the probability of the thermal failure is very low. Test results show that the IVR success probability for CPR1000 is about 99% if the Critical Heat Flux (CHF) of CPR1000 is the same as that of AP600. Further works, including the ERVC enhancement design, the CHF test of the RPV outer wall and the recalculation of the IVR success probability for CPR1000, will be performed at detailed phase in the near future.

scholarly journals Perspectives on a Severe Accident Consequences—10 Years after the Fukushima Accident

2021 ◽  
Vol 2 (4) ◽  
pp. 398-411
Author(s):  
Jinho Song
Keyword(s):  
Nuclear Reactor ◽  
Severe Accident ◽  
Accident Analysis ◽  
Fukushima Accident ◽  
The Public ◽  
Analysis Methodology ◽  
International Attention ◽  
Nuclear Reactor Safety ◽  
Term Monitoring

Scientific issues that draw international attention from the public and experts during the last 10 years after the Fukushima accident are discussed. An assessment of current severe accident analysis methodology, impact on the views of nuclear reactor safety, dispute on the safety of fishery products, discharge of radioactive water to the ocean, status of decommissioning, and needs for long-term monitoring of the environment are discussed.

System and Solenoid Valve Interaction Leading to Spurious Opening

2018 ◽  
Author(s):  
Alton Reich
Keyword(s):  
Storage Tank ◽  
Pressure Transducer ◽  
Water Storage ◽  
Solenoid Valve ◽  
Test Sequence ◽  
Pressure System ◽  
Pressurized Water ◽  
Water Storage Tank ◽  
The Common ◽  
Valve Operation

In a pressurized water reactor the high pressure system vent lines from the pressurizer and reactor are routed to a common header that can be emptied to the refueling water storage tank or a drain tank. During plant testing the valves are operated in the following sequence: the pressurizer isolation valve is opened to pressurize the common header, the pressurizer isolation valve is closed, then the drain tank isolation valve is opened. This sequence of valve operation verifies that the valves open and close properly — opening the pressurizer isolation valve allows steam to enter the common header and is verified by pressure indication via a pressure transducer, and opening the drain tank isolation valve decreases the pressure in the common header and verifies that the pressurizer isolation valve closed properly. During this sequence of valve actuation, the other solenoid valves in the system are subject to transient steam pressures. During one test sequence an isolation valve to the refueling water storage tank indicated that it was not closed for a period of several seconds. Since there is only one pressure transducer in the common header, a systemlevel analysis was performed to obtain a more detailed understanding of the transient pressures in the common header, and how that might affect solenoid valve performance.

Effect of Nitrogen Release From Accumulators on PWR LOCA Analysis

2002 ◽  
Author(s):  
Woon-Shing Yeung ◽  
Ramu K. Sundaram
Keyword(s):  
Coming Out ◽  
Computer Code ◽  
Nitrogen Release ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Concentration Difference ◽  
Dissolved Nitrogen ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Release Model

The accumulator in a Pressurized Water Reactor (PWR) is generally pressurized with inert nitrogen cover gas, and the accumulator water will be saturated with nitrogen. Nitrogen released due to system depressurization during a Loss-of-Coolant Accident (LOCA) transient, consists of the nitrogen that is in the gas phase as well as nitrogen coming out of the liquid from a dissolved state. The effect of nitrogen release from the accumulator on the accident sequence is generally not explicitly addressed in typical LOCA analyses. This paper presents an analytical nitrogen release model and its incorporation into the RELAP5/MOD3 computer code. The model predicts the amount of nitrogen release as a function of concentration difference between the actual and equilibrium conditions, and can track its subsequent transport through the downstream reactor coolant system in a LOCA transient. The model is compared to data from discharge tests with a refrigerant type fluid, pressurized with nitrogen. The results demonstrate that the model is able to calculate the release of the dissolved nitrogen as designed. The modified computer code has been applied to analyze the discharge from a typical PWR accumulator. The results are compared to those obtained without the nitrogen release model. The effect of nitrogen release on major system parameters, including accumulator level, accumulator flow rate, and accumulator pressure, is discussed.

Large Break LOCA Safety Injection Sensitivity for a CE/ABB System 80+ PWR

2002 ◽  
Author(s):  
J. Pottorf ◽  
S. M. Bajorek
Keyword(s):  
Technical Specification ◽  
Nuclear Regulatory Commission ◽  
Injection System ◽  
Engineering System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Core Recovery ◽  
Regulatory Commission ◽  
Saturated Boiling

A WCOBRA/TRAC model of an evolutionary pressurized water reactor with direct vessel injection was constructed using publicly available information and a hypothetical double-ended guillotine break of a cold leg pipe was simulated. The model is an approximation of a ABB/Combustion Engineering System 80+ pressurized water reactor (PWR). WCOBRA/TRAC is the thermal-hydraulics code approved by the U.S. Nuclear Regulatory Commission for use in realistic large break LOCA analyses of Westinghouse 3- and 4-loop PWRs and the AP600 passive design. The AP600 design uses direct vessel injection, and the applicability of WCOBRA/TRAC to such designs is supported by comparisons to appropriate test data. This study extends the application of WCOBRA/TRAC to the investigation of the predicted behavior of direct vessel injection in an evolutionary design. A series of large break LOCA simulations were performed assuming a core power of 3914 MWt, and Technical Specification limits of 2.5 on total peaking factor and 1.7 on hot channel enthalpy rise factor. Two cladding temperature peaks were predicted during reflood, one following bottom of core recovery and a second caused by saturated boiling of water in the downcomer. This behavior is consistent with prior WCOBRA/TRAC calculations for some Westinghouse PWRs. The simulation results are described, and the sensitivity to failure assumptions for the safety injection system is presented.

Study of the Ageing Effects on the Lower Head Failure in a PWR Reactor

2020 ◽  
Author(s):  
R. Lo Frano ◽  
S. Paci ◽  
P. Darnowski ◽  
P. Mazgaj
Keyword(s):  
Nuclear Power ◽  
Quantitative Estimation ◽  
Computer Code ◽  
Severe Accident ◽  
Symmetric Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Term Operation ◽  
Ageing Effects ◽  
Lower Head

Abstract The paper studies influence the ageing effects on the failure of a Reactor Pressure Vessel (RPV) during a severe accident with a core meltdown in a Nuclear Power Plant (NPP). The studied plant is a generic high-power Generation III Pressurized Water Reactor (PWR) developed in the frame of the EU NARSIS project. A Total Station Blackout (SBO) accident was simulated with MELCOR 2.2 severe accident integral computer code. Results of the analysis, temperatures in the lower head and pressures in the lower plenum were used as initial and boundary conditions for the Finite Element Method (FEM) simulations. Two FEM models were developed, a simple two-dimensional axis-symmetric model of the lower head to study fundamental phenomena and complex 3D model to include interactions with the RPV and reactor internals. Ageing effects of a lower head were incorporated into the FEM models to investigate its influence onto lower head response. The ageing phenomena are modelled in terms of degraded mechanical material properties as σ(T), E(T). The primary outcome of the study is the quantitative estimation of the influence of ageing process onto the timing of reactor vessel failure. Presented novel methodology and results can have an impact on future consideration about Long-Term Operation (LTO) of NPPs.

Assessment of Radiological Source Term Releases for Potential Severe Accident Scenarios in a PWR SFP

2018 ◽  
Author(s):  
Kwang-Il Ahn ◽  
Jae-Uk Shin
Keyword(s):  
Water Injection ◽  
Severe Accident ◽  
Base Case ◽  
Source Terms ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Makeup Water ◽  
Accident Scenarios ◽  
The Impact

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.

Break Size Effects on CET Response in an Upper Head SBLOCA Transient

2012 ◽  
Author(s):  
S. Gallardo ◽  
A. Querol ◽  
G. Verdú
Keyword(s):  
Pressure Vessel ◽  
Large Scale ◽  
Injection System ◽  
Test Facility ◽  
Energy Agency ◽  
Pressurized Water ◽  
The Core ◽  
Upward Direction ◽  
Water Reactors ◽  
Accident Conditions

In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced. This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS). The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.

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