The Analysis of Severe Accident Induced Steam Generator Tube Rupture and LERF Risk

2012 ◽  
Vol 614-615 ◽  
pp. 626-631
Author(s):  
Chang Hong Peng ◽  
Ying Hao Yang
Keyword(s):  
Steam Generator ◽  
Creep Rupture ◽  
Severe Accident ◽  
Frequency Model ◽  
Early Release ◽  
Accident Management ◽  
Management Guidance ◽  
The Mean ◽  
Accident Conditions ◽  
Thermal Hydraulic Analysis

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with countercurrent natural circulating high temperature gas in the hot leg and SG tubes. The first step performs thermal-hydraulic analysis to predict the creep rupture parameter of the tubes in severe accident. The next step applies the creep rupture models to test the potential for the degraded SG to rupture before the hot leg. Then, the mean of the SG tube rupture probability was applied to estimate large early release frequency in LERF (Large and Early Release Frequency) model, and the overall LERF risk due to the Induced SGTR was calculated. In the final step, implementation of severe accident management guidance (SAMG), such as the RCS depressurization and refilling to SG, is evaluated using PSA approach. It can be found that strategy of RCS depressurization and refilling to SG can mitigate the result of induced SGTR and LERF effectively.

MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

2014 ◽  
Author(s):  
Pavlin P. Groudev ◽  
Antoaneta E. Stefanova ◽  
Petya I. Vryashkova
Keyword(s):  
Parametric Study ◽  
Cold Water ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Fuel Cladding ◽  
Accident Management ◽  
Management Guidance ◽  
Fuel Behavior ◽  
Accident Conditions

This paper presents the results obtained with the MELCOR computer code from a simulation of fuel behavior in case of severe accident for the VVER-1000 reactor core. The examination is focused on investigation the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the first analyses are modeled options for investigation of melt blockage and debris during the relocation. In the performed analyses are investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. For this purposes it have been performed sensitivity analyses for VVER-1000 reactor core with gadolinium fuel type for parametric study the influence of porosity debris bed. The second analyses is focused on investigation of influence of cold water injection on overheated reactor core at different core exit temperatures, based on severe accident management guidance operator actions. For this purpose was simulated the same SBO scenario with injection of cold water by a high pressure pump in cold leg (quenching from the bottom of reactor core) at different core exit temperatures from 1200 °C to 1500 °C. The aim of the analysis is to track the evolution of the main parameters of the simulated accident. The work was performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed analyses continue the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER-1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER-1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations.

Evaluation of Creep Properties of Alloy 690 Steam Generator Tubes at High Temperature Using Tube Specimen

2019 ◽  
Author(s):  
Jongmin Kim ◽  
Woogon Kim ◽  
Minchul Kim
Keyword(s):  
High Temperature ◽  
Steam Generator ◽  
Creep Test ◽  
Rupture Life ◽  
Creep Rupture ◽  
Severe Accident ◽  
High Temperature Creep ◽  
Creep Life ◽  
Temperature Creep ◽  
Alloy 690

Abstract Thermally induced steam generator (SG) tube failures caused by hot gases from a damaged reactor core can result in a containment bypass event and may lead to release of fission products to the environment. A typical severe accident scenario is a station blackout (SBO) with loss of auxiliary feedwater. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690. Based on reported creep data and creep test results of Alloy 690 in this study, creep life extrapolation was carried out using Larson-Miller Parameter (LMP), Orr-Sherby-Dorn (OSD), Manson-Haferd Parameter (MHP), and Wilshire’s approach. And a hyperbolic sine (sinh) function to determine master curves in LMP, OSD and MHP methods was used for improving the creep life estimation of Alloy 690 material.

CFD Simulation of PANDA and MISTRA Cooler Tests of ERCOSAM-SAMARA Project

2014 ◽  
Author(s):  
A. S. Filippov ◽  
S. Y. Grigoryev ◽  
O. V. Tarasov ◽  
T. A. Iudina
Keyword(s):  
Cfd Simulation ◽  
Molar Fraction ◽  
Transient Behavior ◽  
Severe Accident ◽  
Steam Condensation ◽  
Multi Stage ◽  
Accident Management ◽  
Accident Conditions ◽  
Good Agreement ◽  
Three Phases

The ERCOSAM and SAMARA projects (EURATOM (EU) and ROSATOM (Russia)) include a set of multi-stage experiments carried out at different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). The tests sequences are aimed at investigating hydrogen concentration build-up and stratification during a postulated severe accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and passive auto-catalytic recombiners. Each test includes four phases, of which the first three phases simulate the establishment of severe accident conditions in NPP containment (injection of steam and helium (simulator of hydrogen), stratification of the gas mixture). During the fourth phase of the experiment one of the SAMs simulators is activated. All experiments were simulated at Nuclear Safety Institute of the Russian Academy of Science (IBRAE RAN) with FLUENT and, partially, OpenFOAM codes. In this paper the tests with coolers carried out on PANDA and MISTRA facilities are considered. Their simulations required development of a set of models of volumetric and near-wall condensation phenomena. The models were validated vs. already known tests and vs. integrated experiments of ERCOSAM-SAMARA projects. A brief description of the models and the used CFD methods is provided. Then the results of simulations of the four phases of the tests are presented. Some peculiarities of gas motion and helium distribution obtained in the experiments as well as in their simulations are analyzed. These phenomena concern steam condensation and helium redistribution by convective flows due to the cooler activation in the installation. Local ‘pockets’ of helium are formed with a molar fraction larger than the maximum achieved at the first three phases of the experiments. The accounting of initial and boundary conditions along with calibration of the models provided as a whole a good agreement between calculations and experimental data on transient behavior of gas composition in the facility at the first three phases and at the final fourth phase.

Severe Accident Steady-State Modeling of Advanced Nuclear Power Plant Based on ASTEC Code

2017 ◽  
Author(s):  
Wei Song ◽  
Jiaxu Zuo ◽  
Yan Chen ◽  
Chaojun Li ◽  
Peng Zheng
Keyword(s):  
Steady State ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Cooling Systems ◽  
Accident Management ◽  
Management Guidance ◽  
Level 2 ◽  
Reactor Block

Severe accident is an attractive topic today for the nuclear power plant (NPP) safety. In the nuclear safety regulatory work, it is planned to build a full scale severe accident model for the advanced nuclear power plant of China to study the new designs of severe accident prevention and mitigation systems and strategies, and to further deploy the application on the level 2 PSA and severe accident management guidance. This paper firstly introduces the modeling tool, ASTEC, and then presents the progress of modeling work, which is mainly on the steady state modeling and regulation including reactor block, primary and secondary cooling systems, regulation systems etc. Last but not least, the work plan for the future is given.

Local Failure Modes of a Nuclear Reactor Pressure Vessel Nozzle Under Severe Accident Conditions

2002 ◽  
Author(s):  
Young J. Oh ◽  
Kwang J. Jeong ◽  
Byung G. Park ◽  
Il S. Hwang
Keyword(s):  
Weld Metal ◽  
Pressure Vessel ◽  
Failure Modes ◽  
Nuclear Reactor ◽  
Creep Rupture ◽  
Local Failure ◽  
Severe Accident ◽  
Lower Head ◽  
Accident Conditions ◽  
Reactor Pressure

Most past studies for the creep rupture of a nuclear reactor pressure vessel (RPV) lower head under severe accident conditions, have focused on global deformation and rupture modes. Limited efforts were made on local failure modes associated with penetration nozzles as a part of TMI-2 Vessel Investigation Project (TMI-2 VIP) in 1990’s. However, it was based on an excessively simplified shear deformation model. In the present study, the mode of nozzle failures is investigated using data and nozzle materials from Sandia National Laboratory’s Lower Head Failure Experiment (SNL-LHF). Crack-like separations were revealed at the nozzle weld metal to RPV interfaces indicating the importance of normal stress component rather than the shear stress in the creep rupture. Creep rupture tests were conducted for nozzle and weld metal materials, respectively, at various temperature and stress levels. Stress distribution in the nozzle region is calculated using elastic-viscoplastic finite element analysis (FEA) using the measured properties. Calculation results are compared with earlier results based on the pure shear model of TMI-2 VIE It has been concluded from both LHF-4 nozzle examination and FEA that normal stress at the nozzle/lower head interface is the dominant driving force for the local failure with its likelihood significantly greater than previously assumed.

Development of severe accident management guidance (SAMG) for the Canadian CANDU 6 nuclear power plants

2008 ◽  
Vol 238 (4) ◽  
pp. 1093-1099 ◽  
Author(s):  
Thinh Nguyen ◽  
Raj Jaitly ◽  
Keith Dinnie ◽  
Ron Henry ◽  
Don Sinclair ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management ◽  
Management Guidance
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