Preliminary uncertainty quantification of the core degradation models in predicting the Fukushima Daiichi unit 3 severe accident

2021 ◽  
Vol 382 ◽  
pp. 111383
Author(s):  
Matteo D'Onorio ◽  
Alessio Giampaolo ◽  
Gianfranco Caruso ◽  
Fabio Giannetti
Keyword(s):  
Uncertainty Quantification ◽  
Severe Accident ◽  
The Core ◽  
Degradation Models ◽  
Fukushima Daiichi

Analysis of Accident Progression of Fukushima Daiichi NPP by the SAMPSON Code

2014 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hiroaki Suzuki ◽  
Hideo Mizouchi ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Water Injection ◽  
Heat Removal ◽  
Fission Products ◽  
Severe Accident ◽  
Core Material ◽  
Decay Heat ◽  
The Core ◽  
Fukushima Daiichi ◽  
Fire Engine

This paper describes analysis results of the early phase accident progression of the Fukushima Daiichi Nuclear Power Plant (NPP) Unit 1 by the severe accident analysis code SAMPSON. The isolation condensers were the only devices for decay heat removal at Unit 1, but they stopped after the loss of AC and DC powers. Since there were no decay heat removal for about 14 hours after their termination until the start of alternative water injection into the core by the fire engine, the core melt and the reactor pressure vessel (RPV) bottom failure occurred resulting in large amount of fission products release into the environment. The original SAMPSON was improved by adding new modellings for the phenomena which have been deemed specific to the Fukushima Daiichi NPP: (1) deterioration of SRV gaskets and (2) buckling of in-core-monitor housings which caused the early steam leakage from the core into the drywell, and (3) melt of the in-core-monitor housings in the lower plenum of the RPV. The analysis results showed that (1) 55.3% of UO2 of the initial loading and 66.1% of the core material including UO2, zircaloy, steel and control materials had melted down into the pedestal of the drywell, (2) the amount of Hydrogen generated by Zr-H2O reaction was 686 kg, (3) amount of Cs element released from fuels was 61 kg which was 72% of the total Cs element which was included in fuels at the initiation of the accident, and (4) 18.3% of the corium which fell into the pedestal was one large lump and the 81.7% was particulate corium.

Analysis of Melt Progression and Debris Cooling at the Fukushima Daiichi Unit 3 by the SAMPSON code

2014 ◽  
Author(s):  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Nuclear Power ◽  
Advanced Characterization ◽  
Severe Accident ◽  
Control Rod ◽  
The Core ◽  
Realistic Representation ◽  
Fukushima Daiichi ◽  
Core Conditions ◽  
Definition Of ◽  
And Control

The progression of core melting in a nuclear power plant is characterized by various complications due mainly to chemical reactions and interactions occurring between core materials, which modify the main core temperatures, affecting the time when melt is effectively beginning. On top of it the geometry of a Boiling Water Reactor (BWR) is introducing additional challenges to code modelers. Challenges exist because of the necessity to adopt approximations in the core discretization, which might result in incorrect predictions of the actual phenomena. In particular large uncertainties exist for a clear definition of the fuel and control rod interfaces, discretization of the lower core region as well as control rod guide tubes and monitor penetrations in the lower head. The severe accident code SAMPSON, with an advanced characterization of the core and lower plenum region, is able to recreate a more realistic representation of the phenomena occurring during the melt progression. In this work, SAMPSON description of molten fuel relocation and debris spreading/cooling are presented. State of the art simulation by the SAMPSON code of Fukushima Daiichi unit 3 has shown that core melting progression develops until zirconium is contained within the core, triggering large heat release during oxidation. Thereafter core conditions almost stabilize and 35% of the core falls into the lower plenum, where pipe penetration ejection is predicted releasing molten material within the pedestal.

A model for calculating composition and density of the core melt in the water-moderated water-cooled reactor in case of severe accident

2010 ◽  
Vol 57 (13) ◽  
pp. 1099-1111 ◽  
Author(s):  
V. D. Ozrin ◽  
O. V. Tarasov ◽  
V. F. Strizhov ◽  
A. S. Filippov
Keyword(s):  
Severe Accident ◽  
The Core

The Study of CAP1400 Core Melt and Relocation for IVR Analysis

2017 ◽  
Author(s):  
Jiayun Wang ◽  
Wei Lu ◽  
Pei Wen Gu
Keyword(s):  
Complex Structure ◽  
Pool Analysis ◽  
Large Mass ◽  
Severe Accident ◽  
Retention Strategy ◽  
The Core ◽  
Melt Pool ◽  
Element Simulation ◽  
Simulation Based ◽  
Large Heat

IVR (In-Vessel Retention) strategy is designed as the key severe accident mitigation feature for CAP1400. This paper studies the core melt and relocation progression, which is the base of the melt pool analysis and assessment in the plenum. The MAAP and CFD code are used together to obtain the main insights of the phenomena during core melting. The MAAP code is adopted to have an overall understanding of the progress with the lumped calculation, while the CFD code is used as the tool to study the local failure of the complex structure such as shroud and barrel with finite element simulation. Based on the analysis, the core will heat up after uncovered, and the upper region will melt first to form the core melt pool, as there is still water exist in the active fuel region at the time of upper part rods melting, the debris would be refrozen to form crust to block the relocation. As the melt pool increasing, the shroud is melt-through from the corner, and melts would drop to fill the gap volume between the shroud and barrel before relocation to lower plenum. Furthermore, the barrel will be melted later and the debris relocation to the lower plenum from the core sideward. The melts will touch the lower core support plate before water in the plenum depleted, which provides large mass of metal to be melted into the pool, avoiding large heat flux to challenge the RPV in the pool forming stage.

Ablation of a Solid Material by High Temperature Liquid Jet Impingement: An Application to Corium Jet Impingement on a Sfr Core-Catcher

2021 ◽  
Author(s):  
Alexandre Lecoanet ◽  
Michel Gradeck ◽  
Xiaoyang Gaus-Liu ◽  
Thomas Cron ◽  
Beatrix Fluhrer ◽  
...  
Keyword(s):  
High Temperature ◽  
Deep Understanding ◽  
Jet Impingement ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Liquid Jet ◽  
Cavity Formation ◽  
The Core ◽  
Core Catcher ◽  
Temperature Liquid

Abstract This paper deals with ablation of a solid by a high temperature liquid jet. This phenomenon is a key issue to maintain the vessel integrity during the course of a nuclear reactor severe accident with melting of the core. Depending on the course of such an accident, high temperature corium jets might impinge and ablate the vessel material leading to its potential failure. Since Fukushima Daiichi accident, new mitigation measures are under study. As a designed safety feature of a future European SFR, bearing the purpose of quickly draining of the corium out of the core and protecting the reactor vessel against the attack of molten melt, the in-core corium is relocated via discharge tubes to an in-vessel core-catcher has been planned. The core-catcher design to withstand corium jet impingement demands the knowledge of very complex phenomena such as the dynamics of cavity formation and associated heat transfers. Even studied in the past, no complete data are available concerning the variation of jet parameters and solid structure materials. For a deep understanding of this phenomenon, new tests have been performed using both simulant and prototypical jet and core catcher materials. Part of these tests have been done at University of Lorraine using hot liquid water impinging on transparent ice block allowing for the visualizations of the cavity formation. Other tests have been performed in Karlsruhe Institute of Technology using liquid steel impinging on steel block.

scholarly journals Assessment of Severe Accident Management for Small IPWR under an ESBO Scenario

2019 ◽  
Vol 2019 ◽  
pp. 1-10 ◽  
Author(s):  
Hao Yu ◽  
Minjun Peng
Keyword(s):  
High Pressure ◽  
Low Pressure ◽  
Lower Pressure ◽  
Severe Accident ◽  
Maximum Temperature ◽  
Pressurized Water ◽  
The Core ◽  
Severe Accidents ◽  
Accident Management ◽  
Degradation Studies

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.

scholarly journals EVALUATION METRICS APPLIED TO ACCIDENT TOLERANT FUEL CLADDING CONCEPTS FOR VVER REACTORS

2016 ◽  
Vol 4 ◽  
pp. 89 ◽  
Author(s):  
Martin Sevecek ◽  
Mojmir Valach
Keyword(s):  
Severe Accident ◽  
Evaluation Metrics ◽  
Development Phase ◽  
Fuel System ◽  
Fuel Cladding ◽  
High Interest ◽  
Severe Accidents ◽  
Time Period ◽  
Fukushima Daiichi ◽  
Reactor Types

Enhancing the accident tolerance of LWRs became a topic of high interest in many countries after the accidents at Fukushima-Daiichi. Fuel systems that can tolerate a severe accident for a longer time period are referred as Accident Tolerant Fuels (ATF). Development of a new ATF fuel system requires evaluation, characterization and prioritization since many concepts have been investigated during the first development phase. For that reason, evaluation metrics have to be defined, constraints and attributes of each ATF concept have to be studied and finally rating of concepts presented. This paper summarizes evaluation metrics for ATF cladding with a focus on VVER reactor types. Fundamental attributes and evaluation baseline was defined together with illustrative scenarios of severe accidents for modeling purposes and differences between PWR design and VVER design.

Effects of Ruthenium Release in Oxidizing Conditions on a BWR Source Term

2009 ◽  
Author(s):  
Atso Suopaja¨rvi ◽  
Teemu Ka¨rkela¨ ◽  
Ari Auvinen ◽  
Ilona Lindholm
Keyword(s):  
Pressure Vessel ◽  
Source Term ◽  
Reactor Core ◽  
Severe Accident ◽  
Ruthenium Oxides ◽  
The Core ◽  
The Standard Model ◽  
Reactor Building ◽  
Release Model ◽  
Scoping Studies

The release of ruthenium in oxygen-rich conditions from the reactor core during a severe accident may lead to formation of significantly more volatile ruthenium oxides than produced in steam atmosphere. The effect of volatile ruthenium release in a case a reference BWR nuclear plant was studied to get rough-estimates of the effects on the spreading of airborne ruthenium inside the containment and reactor building and the fission product source term. The selected accident scenario starting during shutdown conditions with pressure vessel upper head opened was a LOCA with a break in the bottom of the RPV. The results suggest that there is a remarkable amount of airborne Ru in the containment atmosphere, unlike with the standard MELCOR Ru release model which predicts no airborne Ru at all in the selected case. The total release of ruthenium from the core can be 5000 times the release predicted by the standard model. Based on the performed plant scoping studies it seems reasonable to take the release of volatile ruthenium oxides into account when assessing source terms for plants during shutdown states.

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