scholarly journals A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

2015 ◽  
Vol 2015 ◽  
pp. 1-14 ◽  
Author(s):  
Diego Mandelli ◽  
Steven Prescott ◽  
Curtis Smith ◽  
Andrea Alfonsi ◽  
Cristian Rabiti ◽  
...  
Keyword(s):  
System Analysis ◽  
Test Case ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Margins ◽  
Particle Hydrodynamics ◽  
Station Blackout ◽  
New Generation ◽  
The Impact

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

scholarly journals Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

2020 ◽  
Vol 6 ◽  
pp. 2 ◽  
Author(s):  
Claire Le Gall ◽  
Fabienne Audubert ◽  
Jacques Léchelle ◽  
Yves Pontillon ◽  
Jean-Louis Hazemann
Keyword(s):  
Oxygen Potential ◽  
Fission Products ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Mox Fuel ◽  
Chemical Behaviour ◽  
Accident Conditions ◽  
The Impact

The objective of this work is to experimentally investigate the effect of the oxygen potential on the fuel and FP chemical behaviour in conditions representative of a severe accident. More specifically, the speciation of Cs, Mo and Ba is investigated. These three highly reactive FP are among the most abundant elements produced through 235U and 239Pu thermal fission and may have a significant impact on human health and environmental contamination in case of a light water reactor severe accident. This work has set out to contribute to the following three fields: providing experimental data on Pressurized Water Reactor (PWR) MOX fuel behaviour submitted to severe accident conditions and related FP speciation; going further in the understanding of FP speciation mechanisms at different stages of a severe accident; developing a method to study volatile FP behaviour, involving the investigation of SIMFuel samples manufactured at low temperature through SPS. In this paper, a focus is made on the impact of the oxygen potential towards the interaction between irradiated MOX fuels and the cladding, the interaction between Mo and Ba under oxidizing conditions and the assessment of the oxygen potential during sintering.

An Estimation of the Dynamic Buckling Load for the Spacer Grid of Pressurized Water Reactor Fuel Assembly

2006 ◽  
Vol 326-328 ◽  
pp. 1603-1606 ◽  
Author(s):  
Sang Youn Jeon ◽  
Young Shin Lee
Keyword(s):  
Fuel Assembly ◽  
Dynamic Buckling ◽  
Buckling Load ◽  
Estimation Methods ◽  
Pressurized Water Reactor ◽  
Spacer Grid ◽  
Pressurized Water ◽  
Water Reactor ◽  
Spacer Grids ◽  
The Impact

This study contains an estimation of the dynamic buckling load for the spacer grid of fuel assembly in pressurized water reactor. Three different estimation methods were proposed for the calculation of the dynamic buckling loads of spacer grid. The dynamic impact tests and analyses were performed to evaluate the impact characteristics of the spacer grids and to predict the dynamic buckling load of the full size spacer grid. The estimation results were compared with the test results for the verification of the estimation methods.

Prospect on the LOCA Analysis Method on the Advanced Pressurized Water Reactor in China

2010 ◽  
Author(s):  
Yan Wang ◽  
Xie Heng
Keyword(s):  
Uncertainty Analysis ◽  
System Analysis ◽  
High Efficiency ◽  
Current System ◽  
National Laboratory ◽  
Pressurized Water Reactor ◽  
Economic Competition ◽  
Pressurized Water ◽  
Water Reactor ◽  
Best Estimate

The LOCA analysis for the advanced pressurized water reactor (PWR) is very important and the methods on it are developing. There are two basic approaches for LOCA (loss of coolant accident) licensing at current. One is based on the conservative requirement of Appendix K of 10CFR50.46 of USNRC, and another is the best estimate (BE) analysis methodology which needs strict sensitivity and uncertainty analysis. The results achieved by the best estimate analysis are closer to the reality than those achieved by the conservative methodology, and the realistic BELOCA analysis in nuclear realm becomes an international trend currently although its development still meet lots of challenges. The research and design on AP1000 to be built in China and larger advanced pressurized water reactor (CAP1400 or CAP1700) as one of Chinese national science & technology major project is in progress. The reliable licensing LOCA analysis as one of the most important accident safety analysis is absolutely necessary. There are three ways to get the code applied in licensing accident analysis: the first way is developing code based on the best estimated methodology with strict uncertainty analysis, the second way is to develop new analysis code based on the conservative Appendix K, and the third way is improving the current system analysis code, which had been verified and validated by many cases, to satisfy the requirements of Appendix K. The last one may be the most feasible way for the AP1000 design with high efficiency and economic competition. Some code like RELAP5 has been used for LOCA analysis, and its results showed good agreement with the test data. RELAP is the transient thermal-hydraulic system analysis code developed by Idaho National Laboratory, in which some model and correlations are not consistent with the conservative requirements of Appendix K, so it can not be applied for licensing LOCA analysis and evaluation directly. In this paper the way to develop analysis code for LOCA license is discussed, and some areas in RELAP code needed to be modified for according with Appendix K are also described, which will be helpful for the advanced PWR design and development in China.

scholarly journals Fatigue Strength Evaluation Of Pressurizer Wall Structure In Pressurized Water Reactor

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Roziq Himawan
Keyword(s):  
Fatigue Strength ◽  
Nuclear Power ◽  
System Analysis ◽  
Fatigue Curve ◽  
Failure Criteria ◽  
Wall Structure ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Design Qualification

<p>Fatigue strength evaluations have been performed to the pressurizer component in Pressurized Water Reactor. Fatigue is the main failure mechanism of material during system in operation. Therefore, this evaluation becomes important to be performed since the pressurizer has a very important function in the reactor’s system. Analysis was performed by using Nuclear Power Plant operation data from 40 years operation and base on Miner theory. This analysis covered all stress level experienced by the reactor during the service. To determine the value of fatigue usage factor a, fatigue curve of SA 533 material was applied. Analysis results show that the cumulative fatigue damage during 40 years in operation is 4,23×10<sup>-4</sup>. This value still far enough below failure criteria, which a value is 1. Therefore, the pressurizer design has already fulfilled the design qualification in term of fatigue aspect.</p>

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