Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

2008 ◽  
Vol 238 (1) ◽  
pp. 33-40 ◽  
Author(s):  
Seiya Hagihara ◽  
Noriyuki Miyazaki
Keyword(s):  
Finite Element Analysis ◽  
Finite Element ◽  
Local Heating ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Creep Failure ◽  
Element Analysis ◽  
Light Water ◽  
Water Reactor ◽  
Accident Condition

Finite Element Analysis for Local Creep of a Tube Coolant Piping System in Light Water Reactor Due to Local Heating Under Severe Accident Condition

2002 ◽  
Author(s):  
Seiya Hagihara ◽  
Noriyuki Miyazaki
Keyword(s):  
Finite Element ◽  
High Temperature ◽  
Constitutive Equation ◽  
Creep Behavior ◽  
Creep Deformation ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Pipe Failure ◽  
Light Water ◽  
Water Reactor

During severe accident of a light water reactor (LWR), reactor coolant piping would be damaged when the piping is subjected to high internal pressure and very high temperature due to heat transfer from high-temperature gas and decay heat from wall-deposited fission product (FP), both from degraded core. In such a case, high-temperature fast creep deformation could be the main cause for the pipe failure. For the evaluation of piping integrity during severe accidents, a method to predict such high-temperature fast creep deformation should be developed, using a creep constitutive equation considering tertiary creep behavior which has not been considered well in the pipe failure analyses. In this paper, a creep constitutive equation was developed, which is based on the Kachanov-Ravotnov isotropic damage rule considered the tertiary creep behavior. Japan Atomic Energy Research Institute (JAERI) creep tensile test data for nuclear-grade cold-drawn SUS316 material was used to determine coefficients of the developed constitutive equation. Using the developed constitutive equation, finite element analyses were performed for local creep deformation of coolant piping under two temperature conditions; uniform temperature and temperature gradient. The analyses results indicated the damage variable being integrated following the evolution of creep damage can indicate pipe wall internal damage condition quantitatively. The damage variable was confirmed further to be able to reproduce the observation in JAERI piping failure tests; pipe failure from the wall outside.

Benchmark of an Application of Acoustic FEA for Light Water Reactor Loss of Coolant Accident Acoustic Loads

2010 ◽  
Author(s):  
Daniel V. Sommerville
Keyword(s):  
Pacific Northwest ◽  
General Purpose ◽  
Light Water Reactor ◽  
Element Analysis ◽  
Blow Down ◽  
Light Water ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Jet Pumps

A Recirculation Line Break Loss of Coolant Accident is a design basis event which must be considered for stress analyses of Boiling Water Reactor internal components such as Jet Pumps and Core Shrouds. This event causes acoustic and fluid drag loads on BWR internals. These loads must also be considered for fracture mechanics evaluations performed to assess allowable operating periods for flaws detected during inservice inspections. Acoustic loads methods generally utilized in the past have been 1-D or simplified 2-D models of the domain of interest. In a few cases sophisticated thermal-hydraulic codes are used to obtain the acoustic response to the LOCA event. The present paper describes the results of a benchmark study performed to validate use of acoustic finite elements available in many commercial general purpose finite element analysis software packages. Use of FEA to predict acoustic loading in the BWR is benchmarked against blow down testing performed by Pacific Northwest Laboratory on a simulated light water reactor vessel. The results of the benchmark demonstrate that use of acoustic FEA yields conservative results and can be considered a viable method for BWR LOCA acoustic load predictions.

scholarly journals Water Hammer Caused by Rapid Gas Production in a Severe Accident in a Light Water Reactor

2005 ◽  
Vol 48 (1) ◽  
pp. 48-55 ◽  
Author(s):  
Fujio INASAKA ◽  
Masaki ADACHI ◽  
Kohki SHIOZAKI ◽  
Izuo AYA ◽  
Hideki NARIAI
Keyword(s):  
Water Hammer ◽  
Gas Production ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Light Water ◽  
Water Reactor

scholarly journals Investigation on Aerosol Pool Scrubbing Model During Severe Accidents

2021 ◽  
Vol 9 ◽  
Author(s):  
L. W. He ◽  
Y. X. Li ◽  
Y. Zhou ◽  
S. Chen ◽  
L. L. Tong ◽  
...  
Keyword(s):  
Bubble Size ◽  
Nuclear Power ◽  
Predictive Ability ◽  
Decontamination Factor ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Spent Fuel ◽  
Analysis Model ◽  
Light Water ◽  
Water Reactor

During a nuclear power plant severe accident, discharging gas mixture into the spent-fuel pool is an alternative containment depressurization measurement through which radioactive aerosols can be scrubbed. However, it is necessary to develop a code for analyzing the decontamination factor of aerosol pool scrubbing. This article has established the analysis model considering key aerosol pool scrubbing mechanisms and introduced the Akita bubble size relationship. In addition, a code for evaluating the decontamination factor of aerosol pool scrubbing was established. The Advanced Containment Experiment and Light Water Reactor Advanced Containment Experiment were simulated with the code considering different bubble sizes of the Akita model and MELCOR default value to verify the suitability of the Akita bubble size model for simulating aerosol pool scrubbing. Furthermore, the simulation results were compared with the results analyzed by MELCOR code and COCOSYS code from literature, and equivalent predictive ability was observed. In addition, a sensitivity analysis on bubble size was conducted, and the contribution of different behaviors and mechanisms has been discussed. Finally, the bubble breakup equation was revised and verified with the conditions of the multi-hole bubbler in the Advanced Containment Experiment and Light Water Reactor Advanced Containment Experiment.

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