Simulation of the HDR E11.2 Hydrogen Mixing Experiment With the CONTAIN Code

2002 ◽  
Author(s):  
Aljaz Skerlavaj ◽  
Ivo Kljenak
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Computer Code ◽  
Severe Accident ◽  
Test Facility ◽  
Hydrogen Distribution ◽  
Flow Geometry ◽  
Input Model ◽  
Containment Test ◽  
Accident Conditions

One of the most well-known experiments on atmosphere stratification in a nuclear power plant containment at severe accident conditions is the test E11.2 “Hydrogen distribution in loop flow geometry”, which was performed in the Heissdampf Reaktor containment test facility in Germany. In the present work, the simulation of the test E11.2 with the CONTAIN computer code is presented. An input model consisting of 72 cells and 263 flowpaths was developed. The predicted pressure history and thermal stratification agree relatively well with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was somewhat underestimated.

THAI Multi-Compartment Containment Test Program

2006 ◽  
Author(s):  
T. Kanzleiter ◽  
G. Poss ◽  
F. Funke ◽  
H.-J. Allelein
Keyword(s):  
Fission Product ◽  
Combined Effect ◽  
Severe Accident ◽  
Test Facility ◽  
Test Program ◽  
Broad Variety ◽  
Containment Test ◽  
Accident Conditions ◽  
Selection Of ◽  
Future Test

The THAI experimental programme includes combined-effect investigations on thermal hydraulics, hydrogen, and fission product (iodine and aerosols) behaviour in LWR containments under severe accident conditions. An overview on the experiments performed up to now and on the future test program is presented, in combination with a selection of typical results to illustrate the versatility of the test facility and the broad variety of topics investigated.

Simulation of Aerosol Behavior in a Saturated Atmosphere With the CONTAIN Code

2002 ◽  
Author(s):  
Ivo Kljenak ◽  
Borut Mavko
Keyword(s):  
Nuclear Power ◽  
Saturated Vapor ◽  
Computer Code ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Experimental Facility ◽  
Standard Problem ◽  
Model Aerosol ◽  
Saturated Atmosphere ◽  
Accident Conditions

Experiments on aerosol behavior in an atmosphere containing saturated vapor, which were performed in the KAEVER experimental facility and proposed for the OECD International Standard Problem No. 44, were simulated with the CONTAIN thermal-hydraulic computer code. The purpose of the work was to assess the capability of the CONTAIN code to model aerosol condensation and deposition in a containment of a light-water-reactor nuclear power plant at severe accident conditions. Results of dry and wet aerosol concentrations in the test vessel atmosphere are presented and analyzed.

scholarly journals Radiation-Tolerance Assessment of a Redundant Wireless Device

2018 ◽  
Vol 170 ◽  
pp. 03004
Author(s):  
Q. Huang ◽  
J. Jiang
Keyword(s):  
Nuclear Power ◽  
Evaluation Method ◽  
Radiation Condition ◽  
Severe Accident ◽  
Radiation Environment ◽  
Radiation Tolerance ◽  
Monitoring Device ◽  
Wireless Monitoring ◽  
Study Results ◽  
Accident Conditions

This paper presents a method to evaluate radiation-tolerance without physical tests for a commercial off-the-shelf (COTS)-based monitoring device for high level radiation fields, such as those found in post-accident conditions in a nuclear power plant (NPP). This paper specifically describes the analysis of radiation environment in a severe accident, radiation damages in electronics, and the redundant solution used to prolong the life of the system, as well as the evaluation method for radiation protection and the analysis method of system reliability. As a case study, a wireless monitoring device with redundant and diversified channels is evaluated by using the developed method. The study results and system assessment data show that, under the given radiation condition, performance of the redundant device is more reliable and more robust than those non-redundant devices. The developed redundant wireless monitoring device is therefore able to apply in those conditions (up to 10 M Rad (Si)) during a severe accident in a NPP.

Onset of Gas Blowthrough During Melt Expulsion From a Hole in a RPV: Numerical Analysis Using the SIMMER Code

2006 ◽  
Author(s):  
Frank Kretzschmar
Keyword(s):  
Numerical Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Material ◽  
Test Facility ◽  
Computer Codes ◽  
Direct Containment Heating ◽  
Heating Up

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.

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.

Simulation of Gas Stratification Build-Up in the Containment Under Severe Accident Conditions

2014 ◽  
Author(s):  
Michele Andreani
Keyword(s):  
Structural Integrity ◽  
Computer Code ◽  
Severe Accident ◽  
Test Sequence ◽  
Small Scale ◽  
High Concentration ◽  
Initial Transient ◽  
Loss Of Coolant Accident ◽  
Computer Codes ◽  
Accident Conditions

The presence of hydrogen stratification in a NPP containment in the case of a severe accident is a source of concern, as pockets of the gas in high concentration could lead to a deflagration or detonation risk, which might challenge the containment structural integrity. These issues, as well as the capability of various computer codes to predict the evolution of a representative accident, are addressed in the coordinated projects ERCOSAM of the 7th EURATOM FWP and the project SAMARA sponsored by ROSATOM. The projects aim to establish whether in a test sequence representative of a severe accident in a LWR hydrogen stratification can be established during the initial transient following a loss of coolant accident (LOCA) and whether and how this stratification can be broken down by the operation of Severe Accident Management systems (SAMs): sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Experiments with helium (as simulant of hydrogen) have been performed at “small scale” in TOSQAN (IRSN, Saclay), and “medium scale” in the MISTRA (CEA, Saclay), PANDA (PSI, Villigen) and SPOT ((JSC “Afrikantov OKBM”, Nizhny Novgorod) facilities. The present paper presents the analysis of the initial transient of some tests in the PANDA, TOSQAN and SPOT facilities using the GOTHIC computer code. The work therefore addresses the capability of the code and a relatively coarse mesh to simulate the pressurisation and build-up of steam and helium stratification for conditions representative of a postulated severe accident scenario, properly scaled to the various facilities. The prediction of the pressurisation is excellent, and the position of the gas concentration stratification front at the end of the steam and helium releases is generally well captured.

scholarly journals Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview

Thermo ◽  
2021 ◽  
Vol 1 (2) ◽  
pp. 151-167
Author(s):  
Hai V. Pham ◽  
Masaki Kurata ◽  
Martin Steinbrueck
Keyword(s):  
Silicon Carbide ◽  
High Temperatures ◽  
Nuclear Power ◽  
Steam Oxidation ◽  
Severe Accident ◽  
Light Water Reactors ◽  
Severe Accidents ◽  
Potential Benefits ◽  
Water Reactors ◽  
Accident Conditions

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

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.

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