The Development of Severe Accident Codes at IRSN and Their Application to Support the Safety Assessment of EPR

2006 ◽  
Author(s):  
Cataldo Caroli ◽  
Alexandre Bleyer ◽  
Ahmed Bentaib ◽  
Patrick Chatelard ◽  
Michel Cranga ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Safety Assessment ◽  
Safety Analysis ◽  
Severe Accident ◽  
Navier Stokes ◽  
Complete Analysis ◽  
Primary System ◽  
Hydrogen Distribution ◽  
Pressurized Water ◽  
Fast Running

IRSN uses a two-tier approach for development of codes analysing the course of a hypothetical severe accident (SA) in a Pressurized Water Reactor (PWR): on one hand, the integral code ASTEC, jointly developed by IRSN and GRS, for fast-running and complete analysis of a sequence; on the other hand, detailed codes for best-estimate analysis of some phenomena such as ICARE/CATHARE, MC3D (for steam explosion), CROCO and TONUS. They have been extensively used to support the level 2 Probabilistic Safety Assessment of the 900 MWe PWR and, in general, for the safety analysis of the French PWR. In particular the codes ICARE/CATHARE, CROCO, MEDICIS (module of ASTEC) and TONUS are used to support the safety assessment of the European Pressurized Reactor (EPR). The ICARE/CATHARE code system has been developed for the detailed evaluation of SA consequences in a PWR primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermalhydraulics French code CATHARE2. The CFD code CROCO describes the corium flow in the spreading compartment. Heat transfer to the surrounding atmosphere and to the basemat, leading to the possible formation of an upper and lower crust, basemat ablation and gas sparging through the flow are modelled. CROCO has been validated against a wide experimental basis, including the CORINE, KATS and VULCANO programs. MEDICIS simulates MCCI (Molten-Corium-Concrete-Interaction) using a lumped-parameter approach. Its models are being continuously improved through the interpretation of most MCCI experiments (OECD-CCI, ACE…). The TONUS code has been developed by IRSN in collaboration with CEA for the analysis of the hydrogen risk (both distribution and combustion) in the reactor containment. The analyses carried out to support the EPR safety assessment are based on a CFD formulation. At this purpose a low-Mach number multi-component Navier-Stokes solver is used to analyse the hydrogen distribution. Presence of air, steam and hydrogen is considered as well as turbulence, condensation and heat transfer in the containment walls. Passive autocatalytic recombiners are also modelled. Hydrogen combustion is afterwards analysed solving the compressible Euler equations coupled with combustion models. Examples of on-going applications of these codes to the EPR safety analysis are presented to illustrate their potentialities.

Validation of the Fast-Running In-Vessel Model ASTRID for Predicting the Radioactive Releases to the Containment

2004 ◽  
Author(s):  
Miettinen Jaakko ◽  
Philipp Schmuck
Keyword(s):  
Process Model ◽  
Emergency Situation ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Primary System ◽  
Developmental Assessment ◽  
Drift Flux ◽  
Fast Running ◽  
Data Process ◽  
Problematic Area

The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) process model is used for the faster than real-time prediction of the radioactivity released into the containment and further into the environment in case of an emergency situation in a light water reactor. Combined together with the containment module COCOSYS the model can predict the entire radioactivity release chain from the primary system to the containment and further into the environment. In the paper the ASTRID thermohydraulic module PROCESS is presented shortly. The thermohydraulic part is a fast running solution for the drift-flux based thermohydraulics. In high temperatures the core degradation leading to the melt pool formation in the reactor barrel and reactor vessel lower head is calculated in the in-vessel module RELOMEL. Finally after the reactor vessel wall has been eroded due to the molten corium in the lower plenum, the massive radioactivity release occurs into the containment. But even before this scenario the radioactivity may be transported from the superheated core to the containment by the coolant. The reference plants for the development have been the Westinghouse type 4-loop PWR, the French type 3-loop PWR, The German type 4-loop Konvoi PWR, the Loviisa VVER type PWR, and the Olkiluoto type internal pump BWR. The reference code for the DBA thermal hydraulics has been the SMABRE code. In the developmental assessment the capability of the rough nodalization of ASTRID has been tested against the SMABRE nodalization describing the plants with 50–500 nodes. For the developmental assessment of the in-vessel severe accident the sample cases are calculated with MELCOR. The more thorough validation is based on the internationally known system codes, RELAP5, MELCOR, CATHARE and ATHLET. In the validation the most problematic area is the radioactivity transport into the containment. This part of the validation is done with the integrated code system.

Estimation of Pressure Loads During Ex-Vessel Steam Explosion

2008 ◽  
Author(s):  
Matjazˇ Leskovar
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Phase 1 ◽  
Interaction Process ◽  
Radioactive Material ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central and side melt pours), the cavity water sub-cooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. Special attention was given to melt droplets freezing, which may significantly influence the outcome of the fuel-coolant interaction process. The performed analysis shows that for some ex-vessel steam explosion scenarios much higher pressure loads are predicted than obtained in the OECD program SERENA Phase 1.

Numerical Two-Phase Flows Simulation and Analysis of the Evolution of the Local Hydrogen Concentration in a PWR Nuclear Containment in the Event of a Severe Accident

2016 ◽  
Author(s):  
Alexandre Zanchetti ◽  
Mickael Hassanaly ◽  
Hervé Cordier ◽  
Antonio Sanna ◽  
Namane Mechitoua ◽  
...  
Keyword(s):  
Hydrogen Concentration ◽  
Nuclear Power ◽  
Cfd Simulation ◽  
Severe Accident ◽  
Water Spray ◽  
Hydrogen Distribution ◽  
Two Phase ◽  
Fukushima Accident ◽  
Pressurized Water ◽  
Spray System

The Fukushima accident reminded us of the possible consequences in terms of radiological release that can result from a hydrogen explosion in a nuclear power plant, and, specifically, within the containment of a water cooled reactor building. Some mitigation means against hydrogen hazards exist but performance improvements in numerical tools simulating thermal-hydraulic flows and hydrogen combustion are necessary to allow realistic assessments of severe accident consequences in the containment. In this context, EDF works on CFD simulation of hydrogen distribution in penalized conditions. After dealing with cases for which the water spray system was assumed to be unavailable, and so treated with single-phase CFD code [1] [2], the present paper content is now about simulation and analysis of the local hydrogen concentration in the case of a severe accident for which the water spray system is available. Numerical developments of a multi-phase CFD code (Neptune_CFD) and code validation lead to consistent simulations. The numerical simulation performed by EDF confirms the favorable safety impact of water spray on pressure and temperature for a LOCA scenario occurring on a 1300 MWe Pressurized Water Reactor. Nevertheless, CFD results show that the activation of the spray system before hydrogen injection gives greater hydrogen concentration. So, in the future, to better assess hydrogen risk, EDF will perform computations at CFD taking into account the interaction between combustion and water sprays.

Pressure Load Estimation During Ex-Vessel Steam Explosion

2009 ◽  
Vol 131 (3) ◽  
Author(s):  
Matjaž Leskovar
Keyword(s):  
Heat Transfer ◽  
Steam Explosion ◽  
Phase 1 ◽  
Interaction Process ◽  
Radioactive Material ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Reactor Accident ◽  
Pressurized Water ◽  
Water Reactor

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor pressure vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel-coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed by varying the location of the melt release (central and side melt pours), the cavity water subcooling, the primary system overpressure at vessel failure, and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. Special attention was given to melt droplet freezing, which may significantly influence the outcome of the fuel-coolant interaction process. The performed analysis shows that for some ex-vessel steam explosion scenarios much higher pressure loads are predicted than obtained in the OECD program SERENA Phase 1.

Evaluation of Advanced Thermal-Hydraulic System Codes for Safety Analysis of Integral Type PWR

2013 ◽  
Author(s):  
J. Choi ◽  
B. Woods
Keyword(s):  
Experimental Data ◽  
Boundary Conditions ◽  
Hydraulic System ◽  
System Analysis ◽  
Natural Circulation ◽  
Safety Analysis ◽  
Test Facility ◽  
Primary System ◽  
Pressurized Water ◽  
Water Reactor

The integral Pressurized Water Reactor (PWR) concept, which contains the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high possibility for near-term deployment. An IAEA International Collaborative Standard Problem (ICSP) on “Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Primary System and Containment during Accidents” has been conducted since 2010. Oregon State University of USA has offered their experimental facility, which was built to demonstrate the feasibility of Multi-Application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven IAEA Member States have been participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiment. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena including the coupling of primary system, high pressure containment and cooling pool are expected to occur in this transient. The ICSP has been conducted in three phases: pre-test (with designed initial & boundary conditions before the conduction of the experiment), blind (with real initial & boundary conditions after the conduction of the experiment) and open simulation (after the observation of real experimental data). Most advanced thermal-hydraulic system analysis codes like TRACE, RELAP5-3D and MARS have been assessed against experiments conducted at MASLWR test facility.

Kinetic Study of Sodium-Water Reaction Phenomena by Differential Thermal Analysis

2012 ◽  
Author(s):  
Shin Kikuchi ◽  
Hiroshi Seino ◽  
Akikazu Kurihara ◽  
Hiroyuki Ohshima
Keyword(s):  
Heat Transfer ◽  
Thermal Analysis ◽  
Differential Thermal Analysis ◽  
Kinetic Study ◽  
Rate Constant ◽  
Reaction Temperature ◽  
Safety Assessment ◽  
Liquid Sodium ◽  
Pressurized Water ◽  
Exponential Factor

In a sodium-cooled fast reactor (SFR), if a heat transfer tube in the steam generator (SG) is failed, high pressurized water vapor blows into the liquid sodium and sodium-water reaction (SWR) takes place. SWR may cause damage to the surface of the neighboring heat transfer tubes by thermal and chemical effects. Therefore, it is important to clearly understand the SWR for safety assessment of the SG. From recent study, sodium (Na)–sodium hydroxide (NaOH) reaction as secondary surface reaction of the SWR phenomena in a SFR was identified by ab initio method [1]. However, kinetics of this reaction is a still open question. It is important to obtain quantitative rate constant of sodium monoxide (Na2O) generation by Na-NaOH reaction because Na2O may accelerate the corrosive and erosive effects. Differential thermal analysis (DTA) provides us with the valuable information on the kinetic parameters, including activation energy, pre-exponential factor (frequency factor) and reaction rate constant. Thus, kinetic study of Na–NaOH reaction has been carried out by using DTA technique. The parameters, including melting points of Na and NaOH, phase transition temperature of NaOH, Na-NaOH reaction temperature and the decomposition temperature of sodium hydride (NaH) were identified from DTA curves. Na, NaOH, and Na2O as major chemical species were observed from the X-ray diffraction (XRD) analysis of the residues after the DTA experiment. It was inferred that Na2O could be generated as a reaction product. Based on the measured reaction temperature, the first-order rate constant of Na2O generation was obtained by the application of the laws of chemical kinetics. From the estimated rate constant, it was found that Na2O generation should be considered during SWR. The results can be the basis for developing a chemical reaction model used in a multi-dimensional sodium-water reaction code, SERAPHIM, being developed by the Japan Atomic Energy Agency (JAEA) toward the safety assessment of the SG in a SFR.

Assessment of ICARE/CATHARE V1 Severe Accident Code

2006 ◽  
Author(s):  
Patrick Chatelard ◽  
Joe¨lle Fleurot ◽  
Olivier Marchand ◽  
Patrick Drai
Keyword(s):  
Safety Assessment ◽  
Severe Accident ◽  
Primary System ◽  
Front End ◽  
Lower Head ◽  
Detailed Evaluation ◽  
Correct Account ◽  
Recommended Guidelines ◽  
Regression Tests ◽  
Level 2

The ICARE/CATHARE code system has been developed by the French “Institut de Radioprotection et de Suˆrete´ Nucle´aire” (IRSN) in the last decade for the detailed evaluation of Severe Accident (SA) consequences in a primary system. It is composed of the coupling of the core degradation IRSN code ICARE2 and of the thermalhydraulics French code CATHARE2. It has been extensively used to support the level 2 Probabilistic Safety Assessment (PSA-2) of the 900 MWe PWR. This paper presents the synthesis of the ICARE/CATHARE V1 assessment which was conducted in the frame of the “International ICARE/CATHARE Users’ Club”, under the management of IRSN. The ICARE/CATHARE V1 validation matrix is composed of more than 60 experiments, distributed in few thermal-hydraulics non-regression tests (to handle the front end phase of a severe accident), numerous Separate-Effect Tests, about 30 Integral Tests covering both the early and the late degradation phases, as well as a “circuit” experiment including hydraulics loops. Finally, the simulation of the TMI-2 accident was also added to assess the code against real conditions. This validation task was aimed at assessing the ICARE/CATHARE V1 capabilities (including the standalone ICARE2 V3mod1 version) and also at proposing recommendations for an optimal use of this version (“Users’ Guidelines”). Thus, with a correct account for the recommended guidelines, it appeared that the last ICARE/CATHARE V1 version could be reasonably used to perform best-estimate reactor studies up to a large corium slumping into the lower head.

MELCOR/VISOR PWR Desktop Simulator for Accident Analysis and Training

2009 ◽  
Author(s):  
Anka de With ◽  
Pieter H. Wakker ◽  
Marcel L. F. Slootman
Keyword(s):  
User Interface ◽  
Graphical User Interface ◽  
Realistic Model ◽  
Severe Accident ◽  
Accident Analysis ◽  
Regulatory Body ◽  
Primary System ◽  
Secondary System ◽  
Pressurized Water ◽  
Interactive Mode

A desktop simulator of the Pressurized Water Reactor (PWR) for training of the accident response has been created under the contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a simulation of the two-loop PWR. It uses the nuclear accident analysis code MELCOR as a computational tool and the visualization tool VISOR as a graphical user interface. The simulator is a realistic model of the plant, mimicking the behaviour of the plant including control and safety systems, and response to user’s actions. Its graphical user interface is user friendly and it displays the coloured mimic of the primary system, containment and the secondary system. The status of important parameters and colour-coded values are displayed on the screen as a part of the plant mimic. Furthermore, parameters of interest can also be shown as a function of time. It can simulate a steady-state run of the power plant and also a variety of accident conditions in auto or interactive mode allowing the user to interfere with the transient, change conditions and analyse the response of the simulator to the interference. The simulator will be used for training of the KFD staff in order to increase the insight in accident progressions. The training will include the use of measures and strategies applied in the Emergency Operating Procedures (EOPs) and the Severe Accident Management Guidelines (SAMGs).

Assessment for Effect of RCS Depressurization and Reactor Cavity Flooding Using PSA Approach

2010 ◽  
Author(s):  
Changhong Peng ◽  
Ning Zhang ◽  
Pingping Liu
Keyword(s):  
Management Strategy ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Systematic Approach ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
The Past ◽  
Accident Management ◽  
Management Guidance

Probabilistic safety assessment (PSA) uses a systematic approach to estimate the reliability and risk of a nuclear power plant (NPP). Over the past few years, severe accident management guidance (SAMG), which delineates the mitigation actions of core melt accidents of an NPP, has been developed to support operators and staff in the technical support center in dealing with those misfortunes. It can be expected that the implementation of SAMG will reduce the amount of radionuclides released to the environment during the accident. The plant studied is a three-loop pressurized water reactor (PWR) with large dry containment. The RCS depressurization and reactor cavity flooding can be used as an accident management strategy. Then, the decrease of LERF (Large and Early Release Frequency) is quantified using PSA approach. It can be found that strategy of RCS depressurization and reactor cavity flooding can mitigate the result of severe accident effectively.

Advanced Model of Radiative Heat Transfer in a Rod Geometry

2007 ◽  
Author(s):  
Gennadii V. Kobelev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev
Keyword(s):  
Heat Transfer ◽  
Heat Exchange ◽  
Radiative Heat Transfer ◽  
Radiative Heat ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Radiative Heat Exchange ◽  
Pressurized Water ◽  
Vessel Elements ◽  
View Factors

Radiative heat transfer is very important in different fields of mechanical engineering and related technologies including heat transfer in furnaces, aerospace, nuclear reactors, different high-temperature assemblies. In particular, in the course of a hypothetical severe accident at pressurized water reactor (PWR) the temperatures inside the reactor vessel reach high values at which taking into account of radiative heat exchange between the structures of reactor (including core and other reactor vessel elements) gets important. Existing models of radiative heat exchange use many limitations and approximations like approximate estimation of view factors and beam lengths. The module MRAD was used in this paper to model the radiative heat exchange in rod-like geometry typical of PWR. Radiative heat exchange is computed using dividing on zones (zonal method) as in existing radiation models implemented to severe accident numerical codes such as ICARE, SCDAP/RELAP, MELCOR but improved in following aspects: • new approach to evaluation of view factors and mean beam length; • detailed evaluation of gas absorptivity and emissivity; • account of effective radiative thermal conductivity for the large core; • account of geometry modification in the course of severe accident. Special attention is paid to deriving of exact analytical values of view factors and mean beam lengths (which are a good tool in radiative heat transfer concerning gas media) for a number of “standard” geometries. Generalized Hottel’s method of strings is used for rods of finite lengths. Monte-Carlo method is used for validation of new model in application to “standard” geometries. The developed model is successfully applied for modeling of PARAMETER-SF1 and QUENCH-06 tests, which use the triangular and square rod assembly respectively.

Sign in / Sign up

Export Citation Format

Share Document