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.

Research on the Loading Pattern of Reactor Core With MOX Fuel in the First Cycle

2012 ◽  
Author(s):  
Tianqi Zhang ◽  
Shihe Yu ◽  
Xinrong Cao
Keyword(s):  
Cross Section ◽  
Reactor Core ◽  
Fuel Utilization ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Loading Pattern ◽  
Mox Fuel ◽  
Initial Cycle ◽  
Calculation Code

In order to research the performance of Pressurized Water Reactor (PWR) with 1/3 MOX fuel in the initial cycle, this paper serves Qinshan II reactor core as the reference core to design suitable MOX assemblies and study relevant core properties. The analyses documented within use assembly cross section calculation code CASMO-4 and core calculation code SIMULATE-3 studied by Studsvik. The purpose of this paper is to demonstrate that the Qinshan II reactor is capable of complying with the requirement for MOX fuel utilization without significant changes to the design of the plant. Several impacts on key physics parameters and safety analysis assumptions, introduced by MOX, are discussing in the paper.

scholarly journals Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2

2021 ◽  
Vol 13 (14) ◽  
pp. 7964
Author(s):  
Alain Flores y Flores ◽  
Danilo Ferretto ◽  
Tereza Marková ◽  
Guido Mazzini
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Nuclear Industry ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Accident Simulation ◽  
Experimental Facilities ◽  
The Difference

The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects.

Numerical Investigation on the Gas Filling Process of Hydrogen-Air-Steam Mixtures in a Semi-Confined Experimental Pipe

2017 ◽  
Author(s):  
Min Li ◽  
Youyou Xu ◽  
Xiaojian Wen ◽  
Songlin Liu ◽  
Guangnan Luo
Keyword(s):  
Severe Accident ◽  
Blockage Ratio ◽  
Gas Composition ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Research ◽  
Filling Time ◽  
Time Required ◽  
Hydrogen Safety

Hydrogen gathering in the containment may occur followed by a severe accident in a nuclear power plant. A flammable mixture can be formed when hydrogen is mixed with air. The ignition of the gas mixture could threaten the integrity of the containment. In order to provide technology base and experiment data for optimization of hydrogen safety technology of Chinese advanced pressurized water reactor CAP1400, a major project regarding hydrogen safety research of pressurized water reactor containment is underway. As an important part of the project, an experimental facility (A4Q-DH) for the study of hydrogen combustion will be built. Gas displacement method is used to filling the premixed hydrogen-air-steam mixtures into the experimental pipe. Flow behaviors of the gases in the pipe are complicated because fluid flow can be disturbed by the built-in obstacles and gas density can be changed with variation of gas composition concentration. Therefore, it is necessary to evaluate the effectiveness of the gas filling method. In this paper, gas filling processes for the experimental pipe with different obstacles and gas composition concentrations were simulated using computational fluid dynamics software ANSYS Fluent. The results indicated that hydrogen-air-steam mixtures can be uniformly distributed in the experimental pipe within tens of seconds. The obstacles with modest blockage ratio in the pipe are conducive to shorten the required gas filling time. The hindering effect of annular obstacles is greater than the one of circular and square obstacles. The time required for air to achieve uniformly distribution increases with the increase of the inlet concentration of steam and hydrogen. However, the time required for hydrogen and steam to be evenly distributed in the pipe are relatively close regardless of the shape and blockage ratio of obstacle and the inlet gas concentration.

Study of Equipment Survivability Under Severe Accident Conditions

2016 ◽  
Author(s):  
Jinquan Yan ◽  
Shanhu Xue ◽  
Lin Tian ◽  
Wei Lu
Keyword(s):  
Nuclear Power ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
New Development ◽  
Nuclear Power Plant Safety ◽  
Accident Conditions ◽  
Accident Sequences

To improve nuclear power plant safety, severe accident prevention and mitigation for both new development and existing plants are generally required by various nuclear safety authorities worldwide. Although great efforts have been made, how to ensure equipment survivability under severe accident conditions is still a concern. This paper depicts an approach to demonstrate the equipment survivability under severe accident conditions by taking passive pressurized water reactor CAP1400 as an instance, including screening of severe accident sequences, determination of bounding environment conditions within containment, equipments identification used for severe accident mitigation and proposed test plan.

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