scholarly journals Probabilistic Dynamics for Integrated Analysis of Accident Sequences considering Uncertain Events

2015 ◽  
Vol 2015 ◽  
pp. 1-18 ◽  
Author(s):  
Robertas Alzbutas
Keyword(s):  
Uncertainty Analysis ◽  
Probabilistic Methods ◽  
Severe Accident ◽  
Integrated Analysis ◽  
Test Case ◽  
Physical Processes ◽  
Dynamic Reliability ◽  
Practical Applications ◽  
Deterministic Modelling ◽  
Accident Sequences

The analytical/deterministic modelling and simulation/probabilistic methods are used separately as a rule in order to analyse the physical processes and random or uncertain events. However, in the currently used probabilistic safety assessment this is an issue. The lack of treatment of dynamic interactions between the physical processes on one hand and random events on the other hand causes the limited assessment. In general, there are a lot of mathematical modelling theories, which can be used separately or integrated in order to extend possibilities of modelling and analysis. The Theory of Probabilistic Dynamics (TPD) and its augmented version based on the concept of stimulus and delay are introduced for the dynamic reliability modelling and the simulation of accidents in hybrid (continuous-discrete) systems considering uncertain events. An approach of non-Markovian simulation and uncertainty analysis is discussed in order to adapt the Stimulus-Driven TPD for practical applications. The developed approach and related methods are used as a basis for a test case simulation in view of various methods applications for severe accident scenario simulation and uncertainty analysis. For this and for wider analysis of accident sequences the initial test case specification is then extended and discussed. Finally, it is concluded that enhancing the modelling of stimulated dynamics with uncertainty and sensitivity analysis allows the detailed simulation of complex system characteristics and representation of their uncertainty. The developed approach of accident modelling and analysis can be efficiently used to estimate the reliability of hybrid systems and at the same time to analyze and possibly decrease the uncertainty of this estimate.

Source Term Analysis Considering B4C/Steel Interaction and Oxidation During Severe Accidents

2017 ◽  
Author(s):  
Jun Ishikawa ◽  
Tomoyuki Sugiyama ◽  
Yu Maruyama
Keyword(s):  
Source Term ◽  
Severe Accident ◽  
Water Pool ◽  
Energy Agency ◽  
Severe Accidents ◽  
Station Blackout ◽  
Carbon Dioxide Co2 ◽  
Accident Sequences ◽  
Term Analysis

The Japan Atomic Energy Agency (JAEA) is pursuing the development and application of the methodologies on fission product (FP) chemistry for source term analysis by using the integrated severe accident analysis code THALES2. In the present study, models for the eutectic interaction of boron carbide (B4C) with steel and the B4C oxidation were incorporated into THALES2 code and applied to the source term analyses for a boiling water reactor (BWR) with Mark-I containment vessel (CV). Two severe accident sequences with drywell (D/W) failure by overpressure initiated by loss of core coolant injection (TQUV sequence) and long-term station blackout (TB sequence) were selected as representative sequences. The analyses indicated that a much larger amount of species from the B4C oxidation was produced in TB sequence than TQUV sequence. More than a half of carbon dioxide (CO2) produced by the B4C oxidation was predicted to dissolve into the water pool of the suppression chamber (S/C), which could largely influence pH of the water pool and consequent formation and release of volatile iodine species.

Code Simulation of Quenching of a High Temperature Debris Bed: Model Improvement and Validation With Experimental Results

2012 ◽  
Author(s):  
A. Bachrata ◽  
F. Fichot ◽  
G. Repetto ◽  
M. Quintard ◽  
J. Fleurot
Keyword(s):  
Late Phase ◽  
Reactor Core ◽  
Injection Rate ◽  
Mitigation Strategies ◽  
Experimental Results ◽  
Severe Accident ◽  
Simulation Tools ◽  
Variation Of Parameters ◽  
Model Improvement ◽  
Accident Sequences

The loss of coolant accidents with core degradation e.g. TMI-2 and Fukushima demonstrated that the nuclear safety analysis has to cover accident sequences involving a late reflood activation in order to develop appropriate and reliable mitigation strategies for both, existing and advanced reactors. The reflood (injection of water) is possible if one or several water sources become available during the accident. In a late phase of accident, no well-defined coolant paths would exist and a large part of the core would resemble to a debris bed e.g. particles with characteristic length-scale: 1 to 5 mm, as observed in TMI-2. The French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation at a late stage of an accident. The purpose of this paper is to propose a consistent thermo-hydraulic model of reflood of severely damaged reactor core for ICARE-CATHARE code. The comparison of the calculations with PRELUDE experimental results is presented. It is shown that the quench front exhibits either a 1D behavior or a 2D one, depending on injection rate or bed characteristics. The PRELUDE data cover a rather large range of variation of parameters for which the developed model appears to be quite predictive.

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.

Enhancing Probabilistic Risk Assessment Methodology for Solving Reactor Safety Issues

2003 ◽  
Author(s):  
F. L. Cho
Keyword(s):  
Nuclear Reactor ◽  
Severe Accident ◽  
Assessment Methodology ◽  
Reactor Safety ◽  
Reactor Accident ◽  
Safety Issues ◽  
Damage States ◽  
Baseline Information ◽  
Risk Assessment Methodology ◽  
Accident Sequences

This paper reveals a paradigm of analyzing the consequential effects of severe nuclear reactor accident, radionuclides fraction and source terms release, that will influence the MACCS2 codification [1], by coupling with the results of SAPHIA-PSA Levels l & 2 quantification process [2], MELCORE [3], STCP [4], PST [5], and XSOR [6]. Those codes are mutually exclusive and useful. However, it lacks of the closed interface and linkage for addressing Plant Damage States (PDS), Severe Accident Sequences, and Risk Consequence. Thus, it is imperative to formulate the consistent baseline information for MACCS2, PSA Levels 1, 2 and 3, and then linking to a new algorithm of NCM.

Reliability Monitor for PWR Safety System Using FMEA and Go-Flow Methodology: Application of Risk Monitor for Nuclear Power Plants

2013 ◽  
Author(s):  
Muhammad Hashim ◽  
Hidekazu Yoshikawa ◽  
Takeshi Matsuoka ◽  
Ming Yang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fault Tree Analysis ◽  
Reliability Evaluation ◽  
Safety System ◽  
Severe Accident ◽  
Monitor System ◽  
Dynamic Reliability ◽  
Defense In Depth ◽  
Risk Monitor

Author’s proposed risk monitor system of Nuclear Power Plant (NPP) is based on the idea of Plant Defense-in-Depth (DiD) risk monitor and reliability monitor to monitor what degree of safety functions incorporated in the plant system is maintained by multiple barriers of Defense-in-Depth (DiD). In the risk monitor system, the range of risk state is not limited in core damage accident but includes all kinds of dangerous states brought by severe accident. In present study, method of the reliability monitor of a risk monitor system is applied to the PWR safety system in order to evaluate the risk state numerically by pursuing all conditions of reliability evaluation given by plant DiD risk monitor. Large break LOCA is taken as an initiating accident event and the implementation of method of the reliability monitor is discussed in detail for single loop PWR safety system by considering the Multilevel Flow Model (MFM), Failure Mode and Effect Analysis (FMEA), and the qualitative reliability evaluation by Fault Tree Analysis (FTA) and the dynamic reliability evaluation by GO-FLOW. The summary of reliability results of PWR safety subsystems are also presented.

Study on the Practical Applications of SVG Technique in New Era

2014 ◽  
Vol 687-691 ◽  
pp. 3481-3484
Author(s):  
Yuan Gao ◽  
Wei Wu Suo ◽  
Xuan Chen Long ◽  
Zhi Ming Yao ◽  
Yi Zhong
Keyword(s):  
Integrated Analysis ◽  
New Era ◽  
Practical Applications ◽  
Reactive Compensation

This paper focused on several practical applications of SVG technique in new era. First, the present study made a basic introduction on reactive compensation from three aspects. Second, it briefly analyzed the features of the load of SVG which remained to be solved in different fields from five aspects. Third, this study made an integrated analysis of the devices and technology parameters of SVG from two aspects. It was expected to be able to provide some useful inspirations and help to the experts and scholars.

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