Integral effect test results on the effect of asymmetric supply of feedwater during a station blackout transient of pressurized water reactor

2020 ◽  
Vol 149 ◽  
pp. 107837
Author(s):  
Kyoung-Ho Kang ◽  
Byoung-Uhn Bae ◽  
Jong-Rok Kim ◽  
Yu-Sun Park ◽  
Jae-Bong Lee ◽  
...  
Keyword(s):  
Test Results ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Integral Effect ◽  
Station Blackout ◽  
Effect Test

Code assessment of ATLAS integral effect test simulating main steam-line break accident of an advanced pressurized water reactor

2017 ◽  
Vol 55 (1) ◽  
pp. 104-112 ◽  
Author(s):  
Kyoung-Ho Kang ◽  
Yu-sun Park ◽  
Byoung-Uhn Bae ◽  
Jong-Rok Kim ◽  
Nam-Hyun Choi ◽  
...  
Keyword(s):  
Steam Line ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Integral Effect ◽  
Main Steam Line ◽  
Main Steam ◽  
Effect Test

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.

A Study on the Static Buckling Load Estimation of the Spacer Grid in the Pressurized Water Reactor Fuel Assembly

2007 ◽  
Vol 353-358 ◽  
pp. 2581-2584 ◽  
Author(s):  
Sang Youn Jeon ◽  
Kyu Tae Kim ◽  
Young Shin Lee
Keyword(s):  
Fuel Assembly ◽  
Estimation Method ◽  
Buckling Load ◽  
Estimation Methods ◽  
Test Results ◽  
Pressurized Water Reactor ◽  
Spacer Grid ◽  
Pressurized Water ◽  
Water Reactor ◽  
Buckling Test

This study contains several estimation methods of the static buckling load for the spacer grid of nuclear fuel assembly in pressurized water reactor. Three different estimation methods were proposed for the calculation of the static buckling loads of spacer grid. The linear and non-linear static buckling analyses were performed to estimate the static buckling load of the spacer grids using ANSYS program. The analyses results were compared with the static buckling test results. Based on the analysis and test results, the applicability of the proposed estimation method for the static buckling load of the spacer grid was investigated.

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