Comparative Study of Physical Models for Particle Sedimentation Using Simmer Code

2020 ◽  
Author(s):  
Eszter Csengeri ◽  
Andrea Bachrata ◽  
Laurent Trotignon ◽  
Elsa Merle
Keyword(s):  
Computer Code ◽  
Mitigation Strategy ◽  
Physical Models ◽  
Severe Accident ◽  
Future Research ◽  
Discharge Process ◽  
Particle Sedimentation ◽  
The Core ◽  
Particulate Solid ◽  
Core Catcher

Abstract In the context of improved safety requirements for Generation IV Sodium-cooled Fast Reactors (SFR), an innovative severe accident mitigation scenario is being investigated. In the French frame of SFR research, the mitigation strategy consists of transfer tubes and a core catcher. The transfer tubes are dedicated to discharge molten fissile materials from the core center region and to guide them towards the core catcher where long-term cooling and sub-critical state may be assured. The physical phenomena occurring during the discharge process are introduced in this paper. The current demonstration of the mitigation strategy uses best-estimate calculations with the reference computer code SIMMER. Previous analyses showed that the material discharge through the transfer tubes might be efficient however, uncertainties of SIMMER approach are identified on the molten material mobility during the relocation process. It is related to a blockage formation due to particulate solid debris accumulation inside the transfer tube, in case of low energy accumulation in the degraded fuel, is believed to originate from the solid particle treatment in the code. As the performance of mitigation strategy strongly depends on the mobility of the relocating mixture, the most predictive behavior of particle flows is of great importance to SFR safety. Therefore, the SIMMER modelling of such flows is analyzed in this work. The first verification and validation test cases regarding the gravitational settling of particle clouds at different volume fractions are presented. Recommendations for reactor calculations and first orientations for future research and development are highlighted.

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.

Numerical Investigation of Heat Transfer Characteristics of Debris Bed After Severe Accident of SFR Based on 1-D Heat Conduction Model

2018 ◽  
Author(s):  
Mengwei Zhang ◽  
Bin Zhang ◽  
Jianqiang Shan
Keyword(s):  
Heat Transfer ◽  
Heat Conduction ◽  
Transfer Process ◽  
Severe Accident ◽  
Heat Transfer Characteristics ◽  
Conduction Model ◽  
One Dimensional ◽  
The Core ◽  
Transfer Characteristics ◽  
Core Catcher

Nuclear reactor severe accidents can lead to the release of a large amount of radioactive material and cause immense disaster to the environment. Since the Fukushima nuclear accident in Japan, the severe accident research has drawn worldwide attention. Based on the one-dimensional heat conduction model, a DEBRIS-HT program for analyzing the heat transfer characteristics of a debris bed after a severe accident of a sodium-cooled fast reactor was developed. The basic idea of the DEBRIS-HT program is to simplify the complex energy transfer process in the debris bed to a simple one-dimensional heat transfer problem by solving the equivalent thermal conductivity in different situations. In this paper, the DEBRIS-HT program code is prepared by using the existing model and compared with the experimental results. The results show that the DEBRIS-HT program can correctly predict the heat transfer process in the fragment bed. In addition, the heat transfer characteristics analysis program is also used to model the core catcher of the China fast reactor. Firstly, the dryout heat flux when all of molten core dropped on the core catcher was calculated, which was compared with the result of Lipinski’s zero dimensional model, and the error between two values is only 11.2%. Then, the temperature distribution was calculated with the heat power of 15MW.

Heat Removal Capability of Core-Catcher in Consideration of Transformation of Cooling Channel

2014 ◽  
Author(s):  
Tomohisa Kurita ◽  
Mitsuo Komuro ◽  
Ryo Suzuki ◽  
Masato Yamada ◽  
Mika Tahara ◽  
...  
Keyword(s):  
Flow Direction ◽  
Natural Circulation ◽  
Heat Removal ◽  
Nucleate Boiling ◽  
Severe Accident ◽  
Flow Section ◽  
Cooling Channel ◽  
Flow Area ◽  
The Core ◽  
Core Catcher

It is necessary to stabilize high temperature molten core in a severe accident for long time without electrical power. The core-catcher is to be installed at the bottom of the lower drywell in order to settle the molten core flowing down from a reactor vessel. Toshiba’s core-catcher system consists of a round basin made up of inclined cooling channels to get natural circulation of the flooding water. So it can cover all pedestal floor and can work in passive manner. We have been confirming an applicability of the core-catcher to actual plants. We have conducted full scaled tests with a unique cooling channel which has inclined rectangular flow section and changing the section area along flow direction in several conditions to evaluate the influence of the parameters on the natural circulation and heat removal capability. The test results showed good heat removal performance with nucleate boiling. However, we should consider a transformation of the cooling channel, for example, by the falling corium. So we calculate the assumed transformation of the cooling channel and conduct natural circulation tests with obstruction in the cooling channel. We confirm that natural circulation flow is stably continues and the cooling channel can remove prescribed heat, even if a flow area have got narrow locally.

scholarly journals Numerical Investigation of Corium Coolability in Core Catcher: Sensitivity to Modeling Parameters

2018 ◽  
Author(s):  
Liancheng Guo ◽  
Andrei Rineiski
Keyword(s):  
Fast Reactor ◽  
Large Scale ◽  
Mass Transfer Process ◽  
Severe Accident ◽  
Liquid Sodium ◽  
Numerical Experience ◽  
The Core ◽  
Sodium Fast Reactor ◽  
Core Catcher ◽  
Fuel Pin

To avoid settling of molten materials directly on the vessel wall in severe accident sequences, the implementation of a ‘core catcher’ device in the lower plenum of sodium fast reactor designs is considered. The device is to collect, retain and cool the debris, created when the corium falls down and accumulates in the core catcher, while interacting with surrounding coolant. This Fuel-Coolant Interaction (FCI) leads to a potentially energetic heat and mass transfer process which may threaten the vessel integrity. For simulations of severe accidents, including FCI, the SIMMER code family is employed at KIT. SIMMER-III and SIMMER-IV are advanced tools for the core disruptive accidents (CDA) analysis of liquid-metal fast reactors (LMFRs) and other GEN-IV systems. They are 2D/3D multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics codes coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. However, the experience of SIMMER application to simulation of corium relocation and related FCI is limited. It should be mentioned that the SIMMER code was not firstly developed for the FCI simulation. However, the related models show its basic capability in such complicate multiphase phenomena. The objective of the study was to preliminarily apply this code in a large-scale simulation. An in-vessel model based on European Sodium Fast Reactor (ESFR) was established and calculated by the SIMMER code. In addition, a sensitivity analysis on some modeling parameters is also conducted to examine their impacts. The characteristics of the debris in the core catcher region, such as debris mass and composition are compared. Besides that, the pressure history in this region, the mass of generated sodium vapor and average temperature of liquid sodium, which can be considered as FCI quantitative parameters, are also discussed. It is expected that the present study can provide some numerical experience of the SIMMER code in plant-scale corium relocation and related FCI simulation.

Flow Characterization and Heat Transfer of Core-Catcher in Passive Safety System

2012 ◽  
Author(s):  
Tomohisa Kurita ◽  
Toshimi Tobimatsu ◽  
Mika Tahara ◽  
Masato Yamada ◽  
Yoshihiro Kojima
Keyword(s):  
Natural Circulation ◽  
Cooling Water ◽  
Heat Removal ◽  
Severe Accident ◽  
Passive Safety ◽  
Cooling Channels ◽  
The Core ◽  
Flow Characterization ◽  
Passive Safety System ◽  
Core Catcher

A mitigation system which can keep core melt stable after a severe accident is necessary to a next generation BWR design. Toshiba has been developing a compact core catcher to be placed at the lower drywell in the containment vessel. The cooling water for the core catcher is supplied from the passive flooder and PCCS drain line. After the core catcher is flooded, the molten core would be cooled by both overflooding water and inclined cooling channels, in which water is boiling and natural circulation is established. So the core catcher can operate in passive manner and has no active component inside the containment. This paper summarizes flow dynamics and heat removal capability in an inclined cooling channel of core catcher when cooling water flows by the natural circulation.

Analysis of Severe Accidents in VVER-1000 Reactors Using the Integral Code ASTEC

2009 ◽  
Author(s):  
Polina Tusheva ◽  
Nils Reinke ◽  
Eberhard Altstadt ◽  
Frank Schaefer ◽  
Frank-Peter Weiss ◽  
...  
Keyword(s):  
Power System ◽  
Failure Time ◽  
Late Phase ◽  
Electric Power System ◽  
Physical Models ◽  
Severe Accident ◽  
Feed Water ◽  
The Core ◽  
Lower Head ◽  
Station Blackout

The studies presented are aiming at a detailed investigation of the behaviour of a VVER-1000/V-320 reactor and the containment structures during a postulated severe accident, including the ways and means by which these accidents may be prevented or mitigated. A hypothetical station blackout scenario (loss of the offsite electric power system concurrent with a turbine trip and unavailability of the emergency AC power system), belonging to the typical beyond design basis accidents, has been investigated. Station blackout results in reactor shut down, loss of feed water and trip of all reactor coolant pumps. Continuous evaporation of the secondary side leads to steam generators’ depletion followed by heating up of the core. In case of unavailability of essential safety systems the core will be severely damaged and finally the reactor pressure vessel (RPV) might fail. The analyses are performed using the integral code ASTEC commonly developed by IRSN (Institut de Radioprotection et de Suˆrete´ Nucle´aire) and GRS (Gesellschaft fu¨r Anlagen- und Reaktorsicherheit mbH). Code-to-code comparative analyses for the early thermal-hydraulic phase have been performed with the GRS code ATHLET. A large number of sensitivity calculations have been done regarding the axial core power distribution, heat losses, and RPV lower head modelling. The analyses have shown that, despite the considerable differences in the codes themselves, the calculation results are similar in terms of thermal hydraulic response. There are discrepancies in timings of phenomena, which are within the limitations of the physical models and the applied nodalizations. It was one objective of this investigation to evaluate the Severe Accident Management (SAM) procedures for VVER-1000 reactors, by for instance estimating the time available for taking appropriate decisions and preparing counter-measures. To evaluate the effect of possible operator actions, a SAM procedure (primary side depressurization) is included into the simulation. Without SAMs, the simulation provides plastic rupture of the RPV after approximately 4.3 h, while with SAMs, a prolongation of the vessel failure time is obtained by approximately 90 minutes. Currently, the late phase of the accident is investigated in more detail by comparing the lower head behaviour as simulated by ASTEC with results from dedicated finite element calculations. The work contributes to the reliability of the ASTEC code by means of plant applications.

Materials Interaction Tests to Identify Base and Coating Materials for an Enhanced In-Vessel Core Catcher Design

2004 ◽  
Author(s):  
J. L. Rempe ◽  
D. L. Knudson ◽  
K. G. Condie ◽  
W. D. Swank ◽  
K. Y. Suh ◽  
...  
Keyword(s):  
High Temperature ◽  
Base Material ◽  
Severe Accident ◽  
Oxide Material ◽  
Coating Materials ◽  
Engineering Research ◽  
The Core ◽  
Lower Head ◽  
Core Catcher ◽  
Core Materials

An enhanced in-vessel core catcher is being designed and evaluated as part of a joint United States (U.S.)–Korean International Nuclear Engineering Research Initiative (INERI) investigating methods to insure In-Vessel Retention (IVR) of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulating oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Initial evaluations suggest that a thermally-sprayed oxide material is the most promising candidate insulator coating for a core catcher. As part of the effort to develop an in-vessel core catcher design, a series of high temperature materials interaction tests were conducted for thermal sprayed coatings and base materials with properties deemed most promising. This paper reports results from these materials interactions tests and efforts to optimize parameters for applying the thermal spray coatings.

scholarly journals ESFR SMART PROJECT CONCEPTUAL DESIGN OF IN-VESSEL CORE CATCHER

2021 ◽  
Vol 247 ◽  
pp. 01002
Author(s):  
Joel Guidez ◽  
Antoine Gerschenfeld ◽  
Janos Bodi ◽  
Konstantin Mikityuk ◽  
Francisco Alvarez-Velarde ◽  
...  
Keyword(s):  
Natural Convection ◽  
Fast Reactor ◽  
Passive Control ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Fukushima Accident ◽  
The Core ◽  
Core Catcher ◽  
Thermal Calculations ◽  
Core Meltdown

Even before Fukushima accident occurred, the safety authorities have required that new power plant designs must take into account beyond design-basis accidents including possible core meltdown. Among the mitigation strategies, the corium retention must be ensured, so a core catcher is implemented in the design of the Generation IV Sodium-cooled Fast Reactor. An internal core catcher within the vessel (in-vessel retention) is the option chosen for the European Sodium-cooled Fast Reactor investigated in the H2020 ESFR-SMART project. The new core investigated in ESFR SMART with lower void effect has a better behavior in case of severe accident. The use of passive control rods is also an improvement for prevention of severe accident. Moreover, we have in the ESFR SMART core dedicated tubes for corium discharge that should allow discharging quickly the melted materials and should help to prevent large criticality. Calculations show that after several seconds, these discharge tubes begin to open, and the corium arrives by this preferential way on the core catcher, quicker and in limited quantities at the beginning of the accident. However, the core catcher is designed to be able to retain the whole core meltdown. Its design allows good possibilities of cooling by natural convection of sodium. Some thermal calculations were provided with a multi-layer concept but the global mechanical conception seems difficult. So a one layer core catcher in molybdenum, material compatible with sodium and used on the core catcher of the last SFR, started in 2016: BN 800, is investigated. Explanations are given on the choice of this material proposed for the catcher and used for thermal calculations. With the proposed design, the corium is spread on the core catcher and the residual power of the corium can be dispelled by natural convection by the sodium circulating around and above the core catcher without boiling of sodium if the melted core is less than about 25% of whole core. In case of bigger quantities of melted core, boiling of sodium could appear under the core catcher. Further less conservative calculations would be necessary to better know the limit.

Analysis of Fuel Coolant Interaction in PWR With Experiment and Computer Code

2010 ◽  
Author(s):  
Xi Wang ◽  
Tian-shu Li ◽  
Xi Huang ◽  
Yan-hua Yang
Keyword(s):  
Low Temperature ◽  
Experimental Research ◽  
Pressure Wave ◽  
Steam Explosion ◽  
Computer Code ◽  
Severe Accident ◽  
Experimental Result ◽  
The Core ◽  
Two Stages

Fuel-Coolant Interaction can be divided into two stages including premixing and explosion. In the stage of premixing, the liquid corium flowing from the core fragmented into drops, and the drops may fragmented into fine fragmentations leads to steam explosion in the condition of triggering. In the severe accident, the contact of melt-fuel and coolant may result into steam explosion, which may threaten integrity of containment. The prediction of steam explosion must be based on results of experimental research and on simulations with computer code. In this work, the experimental result of low temperature fuel-coolant interaction is compared with the computing result of FCI-CMFD code MC3D, analyzing the phenomenon of FCI, the diameter of droplet and the pressure wave of explosion. Then, the MC3D is used in the analysis and prediction of fuel coolant interaction in CPR1000 NPP model.

Mitigation of a Core Meltdown Scenario by Means of a Core Catcher Located Inside the Reactor Pressure Vessel

2004 ◽  
Author(s):  
D. Aquaro ◽  
N. Zaccari
Keyword(s):  
Pressure Vessel ◽  
Effective Conductivity ◽  
Severe Accident ◽  
Mechanical Resistance ◽  
Internal Cooling ◽  
Pebble Bed ◽  
Original Solution ◽  
The Core ◽  
Core Catcher ◽  
Core Meltdown

This paper describes an original solution of core catcher to managing the in vessel retention of the Corium in the accidental event of the core meltdown. The solution envisaged intends to verify the possibility of managing the accidental event within the pressure vessel, ensuring that the CORIUM is confined and cooled. The core catcher, elaborated at the DIMNP, is made of a ceramic pebbel bed (Alumina Al2O3) contained in a metallic or Ceramic Matrix Composite (CMC) structure. The paper illustrates a theoretical model to simulate the thermal-mechanical behaviour of the pebble beds under extremely high loads, developed by the authors. This model has been used to design the core catcher and to determine the effective conductivity and the effective stiffness of the pebble bed. These values have been used in order to implement a numerical model of the core catcher. The results of the thermal and mechanical coupled simulation have permitted to determine the maximum time that the core catcher could resist and the mechanical resistance of the core catcher in the case of RPV external or internal cooling. The preliminary analyses performed have emphasised the good performance of pebble bed core catcher in order to mitigate the envisaged severe accident.

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