Evaluation of High Temperature Tensile and Creep Properties of Light Water Reactor Coolant Piping Materials for Severe Accident Analyses

2000 ◽  
Vol 37 (6) ◽  
pp. 518-529
Author(s):  
Yuhei HARADA ◽  
Yu MARUYAMA ◽  
Akio MAEDA ◽  
Eiichi CHINO ◽  
Hiroaki SHIBAZAKI ◽  
...  
Keyword(s):  
High Temperature ◽  
Light Water Reactor ◽  
Severe Accident ◽  
Creep Properties ◽  
Light Water ◽  
Water Reactor ◽  
Reactor Coolant ◽  
High Temperature Tensile

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.

Fully Coupled Simulation of Oxygen and Heat Diffusion for (U,Pu)O2 Fuel in Both Fast-Breeder Reactor and Light-Water Reactor

2015 ◽  
Vol 1 (4) ◽  
Author(s):  
Wenzhong Zhou ◽  
Rong Liu
Keyword(s):  
High Temperature ◽  
Power Distribution ◽  
Heat Diffusion ◽  
Light Water Reactor ◽  
Fast Breeder Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Ratio Change ◽  
Overall Performance ◽  
Fully Coupled

Oxygen redistribution with a high-temperature gradient is an important fuel performance concern in fast-breeder reactor (FBR) and light-water reactor (LWR) (U,Pu)O2 fuel under irradiation, and affects fuels properties, power distribution, and fuel overall performance. This paper studies the burnup dependent oxygen and heat diffusion behavior in a fully coupled way within (U,Pu)O2 FBR and LWR fuels. The temperature change shows relatively larger impact on oxygen to metal (O/M) ratio redistribution rather than O/M ratio change on temperature, whereas O/M ratio redistributions show different trends for FBR and LWR fuels due to their different deviations from the stoichiometry of oxygen under high-temperature environments.

Evaluation of Fatigue Life of Pressurized Water Reactor Internals Considering Light-Water Reactor Coolant Environmental Effect for Aging Management Program

2014 ◽  
Author(s):  
Jianfeng Yang ◽  
Paul O’Brien
Keyword(s):  
Fatigue Life ◽  
Environmental Effect ◽  
Light Water Reactor ◽  
Pressurized Water ◽  
Light Water ◽  
Water Reactor ◽  
Reactor Coolant ◽  
Plant Life ◽  
Fatigue Evaluation ◽  
Reactor Internals

Most of the current operating nuclear power plants in the United States were designed using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, for fatigue design curves. These design curves were developed in the late 1960s and early 1970s. They were often referred to as “air curves” because they were based on tests conducted in laboratory air environments at ambient temperatures. In recent years, laboratory fatigue test data showed that the light-water reactor environment could have significant impact on the fatigue life of carbon and low-alloy steels, austenitic stainless steel, and nickel-chromium-iron (Ni-Cr-Fe) alloys. United States Nuclear Regulatory Commission, Regulatory Guide 1.207 provides a guideline for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors. It recommend following the method developed in NUREG/CR-6909 [3] when designing reactor coolant pressure boundary components. The industry has invested a lot of effort in developing methods and rules for applying environmental fatigue evaluations for ASME Class 1 components and piping. However, the industry experience in applying the environmental fatigue evaluation for reactor core support structures and internal structures has been very limited. During the recent aging management programs, reactor internal component environmental fatigue evaluations for several pressurized water reactors were evaluated. The analyses calculated the cumulative fatigue usage using the recorded plant-specific transient cycles and the projected cycles for 60 years of plant life. The study concludes that the actual fatigue usages of the components are substantially lower than the specified original design conditions. Even assuming the most severe light-water reactor coolant environmental effects, fatigue will not be a concern for 60 years of plant life. The experiences with environmental fatigue evaluation for reactor internals are still very limited. This study shall provide the industry with beneficial information to develop the approaches and rules addressing the environmental effect on the fatigue life of reactor internals.

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
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