An Evaluation of a Direct Corium Cooling Method for the Ex-Vessel Melt Retention

2010 ◽  
Author(s):  
Kwang Soon Ha ◽  
Hwan Yeol Kim ◽  
Jongtae Kim ◽  
Jong Hwa Park
Keyword(s):  
Steam Explosion ◽  
Power Level ◽  
Water Injection ◽  
Conservation Equation ◽  
Severe Accident ◽  
Injection System ◽  
Steam Generation ◽  
The Core ◽  
Cooling Method ◽  
Core Catcher

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation and in the suppression of a steam explosion. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-inches large break loss of coolant accident without safe injection. The corium spreading regime was estimated by an asymptotic calculation. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water injection system was an effective corium cooling method in the ex-vessel core catcher to preclude a possible steam explosion and to suppress the quick release of steam.

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.

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.

scholarly journals A new direction in PWR simplification

2021 ◽  
Vol 7 ◽  
pp. 2
Author(s):  
Nicolaas M. Bonhomme
Keyword(s):  
High Pressure ◽  
Power Level ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Pressure Control ◽  
Severe Accident ◽  
Injection System ◽  
Primary System ◽  
Ambient Conditions ◽  
Gravity Driven

A new approach to PWR simplification is presented, in which a compact Reactor Coolant System (RCS) configuration is introduced, particularly suited for a power level in the range of 600 MWe. Customary PWR primary system components are eliminated to achieve this RCS simplification. For example, RCS pressure control through a “self-pressurization” mode, with core exit at saturation temperature with less than 1% steam, allows elimination of a pressurizer. Also, mechanical control rods are replaced by reactivity control using negative moderator void and temperature coefficient together with variable speed primary pumps, and with an upgrade in the safety boration function. Decay heat removal in shutdown conditions is realized through the secondary side rather than through primary side equipment. The compact RCS can be installed in a small volume, high-pressure containment. The containment is divided into two leak-tight zones separated by a partition plate. Safety equipment installed in one of the two zones will be protected against adverse ambient conditions from leaks or breaks in the other zone. The partition facilitates management of coolant inventory within the RCS and the containment following RCS leaks or breaks. In particular, the safety injection system as commonly known, consisting of accumulators and multiple stages of injection pumps can be discarded and replaced by gravity-driven flooding tanks. Space available around major RCS components is adequate to avoid compromising accessibility during maintenance or in-service inspection operations. In addition, the two-zone, high-pressure containment provides extra margins in severe accident mitigation. Finally, the proposed containment has a much smaller size than customary large dry containments in PWR practice and it can be anticipated that Nuclear Island building size will similarly be reduced.

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.

Development of a Water Injection System Fusible Plug for Severe Accidents at Nuclear Power Plants

2020 ◽  
Author(s):  
Sei Hirano ◽  
Daisuke Hirasawa ◽  
Yoshihisa Kiyotoki ◽  
Keisuke Sakemura ◽  
Keiji Sasaki ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Reliability ◽  
Water Injection ◽  
Severe Accident ◽  
Injection System ◽  
Normal Operation ◽  
Autonomous Operation ◽  
Performance Requirements

Abstract Background: When terminal stage of Severe Accident (SA) with no coolant injection at a nuclear power plant, the equipment that has cooled and solidified through water injection to a molten core that has ex-vessel and fallen outside of the pressure vessel will then be required to operate autonomously by heat detection, without external signals or power (e.g. electricity, air). The fusible plug operation is triggered by fusible alloy which receives heat from molten core and will melt. Because the fusible plug is also the boundary of Suppression Pool (S/P), high reliability is required for sealing performance. It is for that reason that Hitachi GE Nuclear Energy Ltd. (Hitachi-GE) has developed a fusible plug to serve as a device necessary to operate this system. Features of the Fusible Plug: The autonomous operation of the fusible plug is triggered by the melting of a fusible alloy, which is part of the fusible plug. However, the fusible alloy has a remarkably low mechanical strength and therefore is not suitable as a strength member. As such, it is necessary to ensure reliable plug sealing without applying a load to the fusible alloy so as to prevent the fusible plug from malfunctioning during normal operation. Therefore, to reduce the load to be applied to the fusible alloy, Hitachi-GE has developed a fusible plug structure that operates autonomously by detecting the ambient temperature without using the fusible alloy as a strength member. We have performed a verification test using this fusible plug and confirmed that it satisfies the predetermined performance requirements. Future Actions: Hitachi-GE is holding discussions on using the fusible plug at nuclear power plants in Japan. In the future, we plan to expand to the overseas.

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