Assessment of the MELCOR Code Against PHEBUS Experiment FPT-1 Performed in the Frame of ISP-46

2004 ◽  
Author(s):  
J. Birchley
Keyword(s):  
Good Accuracy ◽  
Current Knowledge ◽  
Fission Products ◽  
Main Tool ◽  
Severe Accident ◽  
Standard Problem ◽  
Nuclear Plants ◽  
Whole Plant ◽  
Computer Codes ◽  
Plant Simulation

Calculations of PHEBUS FPT-1 are performed in the frame of CSNI International Standard Problem ISP-46. The objective of ISP-46 is to assess the capability of computer codes to provide an integral simulation of a severe accident in a Pressurised Water Reactor (PWR), from the initial stages of core heat-up to the behaviour of released fission products in the containment. The present calculations are performed using MELCOR, chosen as the main tool for assessment of Swiss nuclear plants by virtue of its whole-plant simulation capability, using modelling practices as similar as possible to those used in plant analyses. The calculations cover the bundle heat-up, degradation, the release, transport and retention of fission products and other materials, and the thermal-hydraulic and aerosol behaviour in the containment. Comparison between a best-estimate case and experiment demonstrates the code’s ability to capture most aspects of the sequence with fair to good accuracy. Uncertainties remain, particularly in regard to core degradation, and the chemistry and transport of fission products. Weaknesses of code models in these areas largely reflect limitations in current knowledge.

scholarly journals Analytical measurements of fission products during a severe nuclear accident

2018 ◽  
Vol 170 ◽  
pp. 08005 ◽  
Author(s):  
D. Doizi ◽  
S. Reymond la Ruinaz ◽  
I. Haykal ◽  
L. Manceron ◽  
A. Perrin ◽  
...  
Keyword(s):  
Real Time ◽  
Nuclear Reactor ◽  
Fission Products ◽  
Analytical Techniques ◽  
Nuclear Accident ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Dose Rates ◽  
Nuclear Plants ◽  
Management Procedures

The Fukushima accident emphasized the fact that ways to monitor in real time the evolution of a nuclear reactor during a severe accident remain to be developed. No fission products were monitored during twelve days; only dose rates were measured, which is not sufficient to carry out an online diagnosis of the event. The first measurements were announced with little reliability for low volatile fission products. In order to improve the safety of nuclear plants and minimize the industrial, ecological and health consequences of a severe accident, it is necessary to develop new reliable measurement systems, operating at the earliest and closest to the emission source of fission products. Through the French program ANR « Projet d’Investissement d’Avenir », the aim of the DECA-PF project (diagnosis of core degradation from fission products measurements) is to monitor in real time the release of the major fission products (krypton, xenon, gaseous forms of iodine and ruthenium) outside the nuclear reactor containment. These products are released at different times during a nuclear accident and at different states of the nuclear core degradation. Thus, monitoring these fission products gives information on the situation inside the containment and helps to apply the Severe Accident Management procedures. Analytical techniques have been proposed and evaluated. The results are discussed here.

Stand-Alone Containment Analysis of the PHÉBUS FPT-1 Test With the ASTEC and the MELCOR Codes

2016 ◽  
Author(s):  
Bruno Gonfiotti ◽  
Sandro Paci
Keyword(s):  
State Of The Art ◽  
External Environment ◽  
Fission Products ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Complex Task ◽  
Primary Coolant ◽  
Control Volumes ◽  
Coolant System

The estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30–40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior.

scholarly journals Computing operation procedures for chemical plants using whole-plant simulation models

2021 ◽  
Vol 114 ◽  
pp. 104878
Author(s):  
Shumpei Kubosawa ◽  
Takashi Onishi ◽  
Yoshimasa Tsuruoka
Keyword(s):  
Simulation Models ◽  
Chemical Plants ◽  
Whole Plant ◽  
Plant Simulation

scholarly journals A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

2018 ◽  
Vol 2018 ◽  
pp. 1-12
Author(s):  
Taeseok Kim ◽  
Wonjun Choi ◽  
Joongoo Jeon ◽  
Nam Kyung Kim ◽  
Hoichul Jung ◽  
...  
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Fission Products ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Base Case ◽  
Conceptual Approach ◽  
Relief Valve ◽  
Radioactive Nuclides

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies are needed to prevent such unfiltered release of the radioactive nuclides during SGTR accidents. To mitigate the consequence of the SGTR accident, this study was conducted to devise a conceptual approach of installing In-Containment Relief Valve (ICRV) from steam generator (SG) to the free space in the containment building and it was simulated by MELCOR code for numerical analysis. Simulation results show that the radioactive nuclides were not released to the environment in the ICRV case. However, the containment pressure increased more than the base case, which is a disadvantage of the ICRV. To minimize the negative effects of the ICRV, the ICRV linked to Reactor Drain Tank (RDT) and cavity flooding was performed. Because the overpressurization of containment is due to heat of ex-vessel corium, only cavity flooding was effective for depressurization. The conceptual design of the ICRV is effective in mitigating the SGTR accident.

A Model of the Fission Products Release from the Melt Pool during Severe Accident in a Liquid Metal Cooled Fast Reactor

2021 ◽  
Vol 68 (2) ◽  
pp. 152-158
Author(s):  
E. V. Usov ◽  
V. I. Chukhno ◽  
I. A. Klimonov ◽  
V. D. Ozrin ◽  
N. A. Mosunova ◽  
...  
Keyword(s):  
Liquid Metal ◽  
Fast Reactor ◽  
Fission Products ◽  
Severe Accident ◽  
Melt Pool

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.

Whole plant simulation study on load rejection transient of CPR1000 employing advanced Mechanical Shim strategy

2020 ◽  
Vol 142 ◽  
pp. 107397
Author(s):  
Shifa Wu ◽  
Xu Yan ◽  
Xinyu Wei ◽  
Fuyu Zhao ◽  
Shripad Revankar
Keyword(s):  
Simulation Study ◽  
Whole Plant ◽  
Plant Simulation
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