Heat Removal Capability of Core-Catcher With Natural Circulation

Author(s):  
Tomohisa Kurita ◽  
Toshimi Tobimatsu ◽  
Mika Tahara ◽  
Kazuyoshi Aoki ◽  
Yoshihiro Kojima

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.

Author(s):  
Tomohisa Kurita ◽  
Toshimi Tobimatsu ◽  
Mika Tahara ◽  
Masato Yamada ◽  
Yoshihiro Kojima

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.


Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake

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.


Author(s):  
Tomohisa Kurita ◽  
Mitsuo Komuro ◽  
Ryo Suzuki ◽  
Masato Yamada ◽  
Mika Tahara ◽  
...  

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.


Author(s):  
Daiki Takeyama ◽  
Chikako Iwaki ◽  
Yoichi Onitsuka ◽  
Mika Tahara

For mitigation of severe accident with core melting, core-catcher has been developed to catch and cool molten core. The core-catcher developed by Toshiba for Advanced Boiling Water Reactor was designed to be installed under Reactor Pressure Vessel (RPV) and catch molten core in basin with thermal-resistant material. It also has structure including risers and downcomers to generate natural circulation flow of cooling water. On the other hand, there is not enough space to install it in the existing plants due to the height of the inclined cooling channels. Then, we have been developing flat core-catcher with flat cooling channels for the existing plants to reduce total height of the structure. Finned channels will be adopted to increase heat transfer rate by increasing heat transfer area. However, the thermal-hydraulics characteristics of such core-catcher has not been clarified due to the specific configuration, that is, the horizontal rectangular finned channel with the heated surface from upper side. This present study investigated the natural circulation characteristics and heat transfer behavior in the horizontal rectangular finned channel by experiment. Pressure drop, natural circulation flow and temperature were measured by changing heat fluxes. The flow was visualized to obtain flow pattern in the finned channel by a high-speed camera. The maximum value of test range of heat flux was 250 kW/m2, which is the value when the total amount of the molten core would be dropped to the core-catcher. The partial simulated test section of finned channel with 3.5 m length was heated from upper side by heaters to simulate the heat flux from the molten core. This length is the same as the inner diameter of RPV pedestal of the existing plants. The natural circulation mass flux increased as the heat flux increased and then two-phase flow pressure loss also increased. Consequently, circulation flow turned to decrease. As a result of the test, when the heat flux was 50 kW/m2, the circulation mass flux got to the maximum value, 230 kg/m2s. Under all the conditions except for the maximum heat flux of 250 kW/m2, the fin surface temperature was around the saturated temperature. At maximum heat flux 250 kW/m2, the temperature got 540 K. However, the structural soundness was maintained because it is lower than melting temperature of the fin. It can be concluded that the flat and high-thermal-conductivity core-catcher has enough cooling performance to catch and stabilize the molten core.


2018 ◽  
Vol 2018 ◽  
pp. 1-11
Author(s):  
Jiarun Mao ◽  
Lei Song ◽  
Yuhao Liu ◽  
Jiming Lin ◽  
Shanfang Huang ◽  
...  

This paper presents capacity of the passive decay heat removal system (DHRS) operated under the natural circulation conditions to remove decay heat inside the main vessel of the Lead-bismuth eutectic cooled Fast Reactor (LFR). The motivation of this research is to improve the inherent safety of the LFR based on the China Accelerator Driven System (ADS) engineering project. Usually the plant is damaged due to the failure of the main pumps and the main heat exchangers under the Station Blackout (SBO). To prevent this accident, we proposed the DHRS based on the diathermic oil cooling for the LFR. The behavior of the DHRS and the plant was simulated using the CFD code STAR CCM+ using LFR with DHRS. The purpose of this analysis is to evaluate the heat exchange capacity of the DHRS and is to provide the reference for structural improvement and experimental design. The results show that the stable natural circulations are established in both the main vessel and the DHRS. During the decay process, the heat exchange power is above the core decay heat power. In addition, in-core decay heat and heat storage inside the main vessel are efficiently removed. All the thermal-hydraulics parameters are within a safe range. Moreover, the highest temperature occurs at the upper surface of the core. A swirl occurs at the corner of the lateral core surface and some improvements should be considered. And the natural circulation driving force can be further increased by reducing the loop resistance or increasing the natural circulation height based on the present design scenario to enhance the heat exchange effect.


Author(s):  
Ki Won Song ◽  
Shripad T. Revankar ◽  
Hyun Sun Park ◽  
Bo Rhee ◽  
Kwang Soon Ha ◽  
...  

The two-phase natural circulation cooling performance of the APR1400 core catcher system is studied utilizing a drift flux flow model developed via scaling analysis and with an air-water experimental facility. Scaling analysis was carried out to identify key parameters, so that model facility could simulates two-phase natural circulation. In the experimental apparatus, instead of steam, air is injected into the top wall of the test channel to simulate bubble formation and void distribution due to boiling water in the core catcher channel. Measurement of void fraction critical to the heat transfer between the wall and coolant is carried out at certain key position using double-sensor conductivity probes. Results from the model provide expected natural circulation flow rate in the cooling channel of the core catcher system. The observed flow regimes and the data on void fraction are presented. For a given design of the down comer piping entrance condition bubble entrainment was observed that significantly reduced the natural circulation flow rate.


Author(s):  
A. K. Nayak ◽  
Mukesh Kumar ◽  
Sumit V. Prasad ◽  
V. Jain ◽  
D. K. Chandraker

Removal of decay heat with nonavailability of active systems is a safety issue especially during station blackout (SBO) in a light water reactor. Passive systems are being incorporated in the new designs of nuclear reactors for this purpose. Some of the advanced reactors such as Indian advanced heavy water reactor (AHWR) have dedicated isolation condensers (ICs) which are submerged in large water pool called gravity driven water pool (GDWP). These ICs remove decay heat from the core by natural circulation cooling and dissipate it to the GDWP by natural convection. There is a concern that cracks may develop in the GDWP if a large seismic event similar to Fukushima type occurs. In that case, the pool water is lost and it can threaten the core coolability because of loss of heat sink. In AHWR, the cracks in the water pool leads to the relocation of the water of the pool to the reactor cavity. Feeders of AHWR are positioned in the reactor cavity. Thus, the water relocated in the cavity, will eventually submerge the feeders and these submerged feeders have the potential to remove the decay heat of the core. However, the feeders are located at a lower elevation as compared to the core, and hence, there is concern on the heat removal capability by the submerged feeders by natural convection. To understand this aspect and to establish the core coolability under the above-mentioned conditions, experiments were performed in a full-scale test facility of AHWR. Experiments showed that the decay heat can be safely removed in natural circulation mode of cooling with heat sink located at lower elevation than the heat source.


Author(s):  
Manfred Fischer

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.


Author(s):  
N. Ueda ◽  
I. Kinoshita ◽  
Y. Nishi ◽  
A. Minato ◽  
H. Matsumiya ◽  
...  

This paper describes the passive safety features utilized in the updated sodium cooled Super-Safe, Small and Simple fast reactor, which is the improved 4S reactor. This reactor can operate up to ten years without refueling and neutron reflector regulates the reactivity. One of the design requirements is to secure the core against all anticipated transients without reactor scram. Therefore, the reactor concept is to design to enhance the passive safety features. All temperature reactivity feedback coefficients including whole core sodium void worth are negative. Also, introducing of RVACS (Reactor Vessel Auxiliary Cooling System) can enhance the passive decay heat removal capability. Safety analyses are carried out to simulate various transient sequences, which are loss of flow events, transient overpower events and loss of heat sink events, in order to evaluate the passive safety capabilities. A calculation tool for plant dynamics analyses for fast reactors has been modified to model the 4S including the unique plant system, which are reflector control system, circulation pumps and RVACS. The analytical results predict that the designed passive features improve the safety in which temperature variation in transients are satisfied with the safety criteria for the fuel element and the structure of the primary coolant boundary.


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