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.

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.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

Heat Removal Capability of Core-Catcher With Natural Circulation

2013 ◽  
Author(s):  
Tomohisa Kurita ◽  
Toshimi Tobimatsu ◽  
Mika Tahara ◽  
Kazuyoshi Aoki ◽  
Yoshihiro Kojima
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Dimensionless Number ◽  
Dimensionless Numbers ◽  
Cooling Channels ◽  
The Core ◽  
Water Chamber ◽  
Core Catcher ◽  
Fundamental Equations

Toshiba has developed a core-catcher system. It is to be installed at the bottom of the lower drywell in order to stabilize a molten core flowing down from a reactor vessel. It consists of a round basin made up of inclined cooling channels arranged axisymmetrically, and the structure including risers, downcomers and a water chamber to get natural circulation of the flooding water. So it can cover entire pedestal floor and can work in passive manner. In order to confirm the heat removal capability of the core catcher with natural circulation, we have conducted full scaled tests in several conditions. Some important dimensionless numbers obtained from fundamental equations of the natural circulation are used for the tests. Using dimensionless number and to compare with several analysis, we can verify that the experiment is adequate to simulate the actual plant.

The Core Melt Stabilization Concept of the EPR and Its Experimental Validation

2006 ◽  
Author(s):  
Manfred Fischer
Keyword(s):  
Molten Pool ◽  
Heat Removal ◽  
Volume Ratio ◽  
Spatial Separation ◽  
Severe Accident ◽  
Design Stage ◽  
Pressurized Water ◽  
The Core ◽  
Core Catcher ◽  
Accident Conditions

The strategy of the European Pressurized Water Reactor (EPR) to avoid severe accident conditions is based on the improved defense-in-depth approaches of the French “N4” and the German “Konvoi” plants. In addition, the EPR takes measures, at the design stage, to drastically limit the consequences of a postulated core-melt accident. The latter requires a strengthening of the confinement function and a significant reduction of the risk of short- and long-term containment failure. Scenarios with potentially high mechanical loads and large early releases like: high-pressure RPV failure, global hydrogen detonation, and energetic steam explosion must be prevented. The remaining low-pressure sequences are mitigated by dedicated measures that include hydrogen recombination, sustained heat removal out of the containment, and the stabilization of the molten core in an ex-vessel core catcher located in a compartment lateral to the pit. The spatial separation protects the core catcher from loads during RPV failure and, vice versa, eliminates concerns related with its unintended flooding during power operation. To make the relocation of the melt into the core catcher scenario-independent and robust against the uncertainties associated with in-vessel molten pool formation and RPV failure, the corium is temporarily retained, accumulated and conditioned in the pit during interaction with a sacrificial concrete layer. Spreading of the accumulated molten pool is initiated by penetrating a concrete plug in the bottom. The increase in surface-to-volume ratio achieved by the spreading process strongly enhances quenching and cool-down of the melt after flooding. The required water is passively drained from the IRWST. After availability of the containment heat removal system the steam from the boiling pool is re-condensed by sprays. The CHRS can also optionally cool the core catcher directly, which, in consequence, establishes a sub-cooled pool near-atmospheric pressure levels in the containment. The described concept rests on a large experimental knowledge base which covers all main phenomena involved, including melt interaction with structural material, melt spreading, melt and quenching, as well as the efficacy of the core catcher cooling. Besides giving an overview of the EPR core melt mitigation concept, the paper summarizes its R&D bases and describes which conclusions have been drawn from the various experimental projects and how these conclusions are used in the validation of the EPR concept.

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.

Analysis and Research on Natural Circulation Capacity of HFETR

2016 ◽  
Author(s):  
Taozhong Xu ◽  
Caiyu Deng ◽  
Yuxin Xiang
Keyword(s):  
Pressure Vessel ◽  
Cold Water ◽  
Flow Direction ◽  
Natural Circulation ◽  
Reactor Power ◽  
Hot Water ◽  
Forced Circulation ◽  
The Core ◽  
Establishment Process ◽  
Residual Heat

Natural circulation is being used as an important circulation to remove reactor residual heat. In the core of High Flux Engineering Trial Reactor of China (HFETR), the coolant is driven by pumps normally and flows from upside to downside in the core. When HFETR is shut down or runs in low power, the natural circulation between the hot water in the core and the cold water in the reflector inside the pressure vessel is established to cool down the core. Since the natural circulation processed only in the pressure vessel, the accident pumps need to be turned on periodically to remove reactor residual heat. The inversion of flow direction in HFETR and internal natural circulation lead to a different natural circulation establishment process from traditional reactor in which coolant flows form down to top normally. In this paper the transition between the natural circulation and forced circulation is analyzed by RELAP5/MOD3 code. The results showed that the accident pump could be turned off in the power of 850kW; The time, at which the accident pump needs to be turned on to transit the natural circulation to forced circulation, is decided by the temperature of the water in top of pressure vessel, and a formula between temperature of the water in the top of pressure vessel and the reactor power was obtained. The research results have theoretical and practical value to the full use of the natural circulation ability, as well as the safety of the engineering reactors or similar test facilities.

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.

Fuzzy Linear Matrix Inequalities Control for a Rectangular Thermosyphon

2012 ◽  
Vol 532-533 ◽  
pp. 496-502
Author(s):  
Kun Qian ◽  
Xiang Ping Pang ◽  
Hao Yan ◽  
Kai Liu
Keyword(s):  
Linear Matrix Inequalities ◽  
Flow Direction ◽  
Fluid Dynamic ◽  
Natural Circulation ◽  
Fuzzy Model ◽  
Heat Removal ◽  
Linear Matrix ◽  
Thermal Source ◽  
Matrix Inequalities ◽  
Temperature Oscillations

This paper proposed a fuzzy Linear Matrix Inequalities(LMI) control strategy for a rectangular thermosyphon. These thermo-fluid-dynamic systems are used to refrigerate heat sources by means of natural circulation of a fluid without mechanical pumping. Two different fuzzy LMI controllers have been designed in order to guarantee closed loop stability, on the basis of a suitable nonlinear T-S fuzzy model of the system. In particular the controllers are designed in order to prevent temperature oscillations, associated with the inversion of the flow direction, which compromise the heat removal from the thermal source. Two different strategies have been adopted to introduce suitable constraints on the control action.

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