Semi-analytical modeling of a PWR lower head failure under severe accident conditions using an axisymetrical shell theory

2005 ◽  
Vol 235 (8) ◽  
pp. 845-853 ◽  
Author(s):  
V. Koundy ◽  
I. Cormeau
Keyword(s):  
Analytical Modeling ◽  
Shell Theory ◽  
Severe Accident ◽  
Lower Head ◽  
Accident Conditions

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.

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

Ruthenium release modelling in air and steam atmospheres under severe accident conditions using the MAAP4 code

2012 ◽  
Vol 246 ◽  
pp. 157-162 ◽  
Author(s):  
Emilie Beuzet ◽  
Jean-Sylvestre Lamy ◽  
Hadrien Perron ◽  
Eric Simoni ◽  
Gérard Ducros
Keyword(s):  
Severe Accident ◽  
Accident Conditions

Study of Refractory Material Installation in the Reactor Cavity

2014 ◽  
Author(s):  
Young Tae Moon ◽  
In Chul Ryu ◽  
Quan Zhou ◽  
Paul McMinn ◽  
Chan Y. Paik
Keyword(s):  
Refractory Material ◽  
Protective Layer ◽  
Ablation Depth ◽  
Severe Accident ◽  
Refractory Materials ◽  
Protective Material ◽  
Concrete Material ◽  
Combustible Gases ◽  
Accident Conditions ◽  
Containment Pressure

During a severe accident with a vessel failure, corium relocates from the vessel into the reactor cavity (PWR) or pedestal (BWR) and accumulates on top of the cavity floor to form a corium pool. This corium pool is hot enough to cause a Molten Corium-Concrete Interaction (MCCI) that can ablate the concrete structure even if water is present on top of the corium. MCCI will also produce steam and other gases that increase containment pressure as well as generate combustible gases (Hydrogen and Carbon Monoxide). Current MAAP5* calculations with conservative assumptions have shown that the ablation depth in a basemat constructed of siliceous concrete can be larger than the depth of liner, even if the reactor cavity is flooded by water. To retain the melt in the containment and to cool the corium pool before the erosion reaches the liner plate, several approaches are being considered. One of these approaches is the installation of a protective layer on top of the concrete floor to retard MCCI. The purpose of this paper is to study the performance of different protective materials under postulated severe accident conditions. The candidates for the protective materials are refractory materials and limestone/limestone-common-sand (LCS) concrete. The refractory material was chosen based on the thermal performance and dissolution rate of the refractory material calculated by analytical calculations and also by MAAP5. Adding the refractory protective material protects the underlying concrete basemat from melting temporarily, so that water ingression into the surface of the corium is not initially affected by addition of the concrete material. *MAAP5 is an integrated severe accident code owned by the Electric Power Research Institute and developed by Fauske and Associates, LLC.

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