The ASTEC Integral Code for Severe Accident Simulation

2009 ◽  
Vol 165 (3) ◽  
pp. 293-307 ◽  
Author(s):  
J. P. van Dorsselaere ◽  
C. Seropian ◽  
P. Chatelard ◽  
F. Jacq ◽  
J. Fleurot ◽  
...  
Keyword(s):  
Severe Accident ◽  
Accident Simulation

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

scholarly journals Codes for NPP severe accident simulation: development, validation and applications

2015 ◽  
Vol 5 (2) ◽  
pp. 7-14
Author(s):  
ARKADIY E. KISELEV
Keyword(s):  
Reactor Core ◽  
General Information ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Software Systems ◽  
Core Catcher ◽  
Accident Simulation ◽  
Integrated Software ◽  
Academy Of Sciences

The software tools that describe various safety aspects of NPP with VVER reactor have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN). Functionally, the codes can be divided into two groups: the calculation codes that describe separate elements of NPP equipment and/or a group of processes and integrated software systems that allow solving the tasks of the NPP safety assessment in coupled formulation. In particular, IBRAE RAN in cooperation with the nuclear industry organizations has developed the integrated software package SOCRAT designed to analyze the behavior of NPP with VVER at various stages of beyond-design-basis accidents, including the stages of reactor core degradation and long-term melt retention in a core catcher. The general information about development, validation and applications of SOCRAT code is presented and discussed in the paper.

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.

S083026 Development of severe accident simulation visualization tool

2012 ◽  
Vol 2012 (0) ◽  
pp. _S083026-1-_S083026-3
Author(s):  
Akio ARAKAWA ◽  
Yutaka TAKEUCHI ◽  
Nobuyuki SAIJOU ◽  
Yasushi YAMAMOTO
Keyword(s):  
Severe Accident ◽  
Visualization Tool ◽  
Accident Simulation

ADAM: An Accident Diagnostic, Analysis and Management System — Applications to Severe Accident Simulation and Management

2002 ◽  
Author(s):  
M. J. Zavisca ◽  
M. Khatib-Rahbar ◽  
H. Esmaili ◽  
R. Schulz
Keyword(s):  
Real Time ◽  
Water Injection ◽  
Management Training ◽  
Computer Applications ◽  
Severe Accident ◽  
Diagnostic Analysis ◽  
Gas Generation ◽  
Accident Management ◽  
Accident Simulation ◽  
On Line

The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM’s severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of “alarms” are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management.

A Method for Online Calibration of MAAP Simulations in a Severe Accident Management System Database

2017 ◽  
Author(s):  
Weifeng Xu ◽  
Fangqing Yang ◽  
Peng Chen ◽  
Yehong Liao
Keyword(s):  
Management System ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Outlet Temperature ◽  
Online Calibration ◽  
Support Plate ◽  
Initial Event ◽  
Accident Management ◽  
Accident Simulation ◽  
Plate Failure

During a nuclear plant accident, five accident events are usually considered, including core uncovery, core outlet temperature arrived at 650 °C, core support plate failure, reactor vessel failure and containment failure. In accident emergency aspect, when an accident happens, the initial event can be utilized in the severe accident management system which is based on MAAP to simulate the long process of the accident, so as to provide support for operators to take actions. However, in MAAP, many sensitivity parameters exist, which reflect phenomenological uncertainty or models uncertainty and will influence the happening time of the five accident events above. In this paper, based on MAAP5 and LOCAs, the CPR1000 is simulated to analyze the influences of MAAP5’s sensitivity parameters reflecting phenomenological uncertainty on the accident process, which is aimed to find out the sensitivity parameters associated to the five important accident events and build the database between these sensitivity parameters and five accident events’ happening time. Then, based on the research above, a preliminary approach to optimize the MAAP5’s accidents simulation is introduced, which is realized by adjusting sensitivity parameters. Finally, the application of this research will be showed in a severe accident management system developed by us. The research results offer great reference significance for the severe accident simulation and prediction in MAAP5.

scholarly journals Severe accident simulation at Olkiuoto

1995 ◽  
Author(s):  
H Tirkkonen ◽  
T Saarenpaeae ◽  
L C Cliff Po
Keyword(s):  
Severe Accident ◽  
Accident Simulation

Introduction of Three Methods Used for the Nuclear Accident Diagnosis

2017 ◽  
Author(s):  
Peng Chen ◽  
Weifeng Xu ◽  
Fangqing Yang ◽  
Yehong Liao
Keyword(s):  
Management System ◽  
Nuclear Power ◽  
Nuclear Accident ◽  
Severe Accident ◽  
Severe Accidents ◽  
Accident Management ◽  
Accident Simulation ◽  
Diagnosis Method ◽  
Nuclear Severe Accidents ◽  
Diagnosis Technique

In order to deal with the nuclear severe accidents, the severe accident management systems are popularly considered and developed at home and abroad recently. A severe accident management system usually includes these functional parts: accident monitor, accident diagnosis, accident simulation, accident prognosis and SAMG support. Here, the accident diagnosis part is mainly concerned, and three nuclear accident diagnosis methods are introduced here, including BP neural network method, SDG expert diagnosis technique and artificial diagnosis method, which are also applied to a severe accident management system developed by us. In this paper, firstly, the severe accident management system developed by us will be introduced briefly. Then, three accident diagnosis methods for nuclear power plant (NPP) are showed and described in detail. At last, two cases including LOCA and SGTR accidents are used for the verification of these accident diagnosis methods and some analyzing results and conclusions are given. The results show that the three diagnosis methods are very useful for the accident diagnosis of NPP, which can diagnose the accident type accurately and offer much information or support to the severe accident management system and operators. The paper offers some reference significance for the research of accident diagnosis methods and the development of severe accident management system.

The Development and Application of PWR Severe Accident Simulation Model

2013 ◽  
Author(s):  
Yanfang Chen ◽  
Zhengquan Xie ◽  
Xusheng Lin ◽  
Fuchang Shan ◽  
Wei Wei ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Mathematical Models ◽  
Simulation Model ◽  
Nuclear Power ◽  
Severe Accident ◽  
Plant Type ◽  
Simulation Code ◽  
Accident Simulation ◽  
Simulation Codes

The severe accident simulation codes that developed by the engineers in RINPO are introduced in this chapter. The results of the severe accident caused by large LOCA plus losing safety core injection are presented. Comparison with the results of SCDAP/RELAP5/MOD3.2 of the same accident and the same nuclear power plant type has been made. From the comparison and the analysis we can make the conclusion that the trend-lines are correct and the mathematical models are reasonable in this simulation code.

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