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

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):  
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):  
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):  
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.


2013 ◽  
Vol 2013 ◽  
pp. 1-8
Author(s):  
Ge Shao ◽  
Lili Tong ◽  
Xuewu Cao

To prevent HPME and DCH, SADV is proposed to be added to the pressurizer for Chinese improved 1000 MWe PWR NPP with the reference of EPR design. Rapid depressurization capability is assessed using the mechanical analytical code. Three typical severe accident sequences of TMLB’, SBLOCA, and LOFW are selected. It shows that with activation of the SADV the RCS pressure is low enough to prevent HPME and DCH. Natural circulation at upper RPV and hot leg is considered for the rapid depressurization capacity analysis. The result shows that natural circulation phenomenon results in heat transfer from the core to the pipes in RCS which may cause the creep rupture of pipes in RCS and delays the severe accident progression. Different SADV valve areas are investigated to the influence of depressurization of RCS. Analysis shows that the introduction of SADV with right valve area will delay progression of core degradation to RPV failure. Valve area is to be optimized since smaller SADV area will reduce its effect and too large valve area will lead to excessive loss of water inventory in RCS and makes core degradation progression to RPV failure faster without additional core cooling water sources.


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.


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.


2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.


Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 244-255
Author(s):  
S. H. Abdel-Latif ◽  
A. M. Refaey

Abstract The AP600 is a Westinghouse Advanced Passive PWR with a two–loop 1 940 MWt. This reactor is equipped with advanced passive safety systems which are designed to operate automatically at desired set-points. On the other hand, the failure or nonavailability to operate of any of the passive safety systems may affect reactor safety. In this study, modeling and nodalization of primary and secondary loops, and all passive reactor cooling systems are conducted and a 10-inch cold leg break LOCA is analyzed using ATHLET 3.1A Code. During loss of coolant accident in which the passive safety system failure or nonavailability are considered, four different scenarios are assumed. Scenario 1 with the availability of all passive systems, scenario 2 is failure of one of the accumulators to activate, scenario 3 is without actuation of the automatic depressurization system (ADS) stages 1–3, and scenario 4 is without actuation of ADS stage 4. Results indicated that the actuation of passive safety systems provide sufficient core cooling and thus could mitigate the accidental consequence of LOCAs. Failure of one accumulator during LOCA causes early actuation of ADS and In-Containment Refueling Water Storage Tank (IRWST). In scenario 3 where the LOCA without ADS stages 1–3 actuations, the depressurization of the primary system is relatively slow and the level of the core coolant drops much earlier than IRWST actuation. In scenario 4 where the accident without ADS stage-4 activation, results in slow depressurization and the level of the core coolant drops earlier than IRWST injection. During the accident process, the core uncovery and fuel heat up did not happen and as a result the safety of AP600 during a 10-in. cold leg MBLOCA was established. The relation between the cladding surface temperature and the primary pressure with the actuation signals of the passive safety systems are compared with that of RELAP5/Mode 3.4 code and a tolerable agreement was obtained.


Author(s):  
Alexandre Lecoanet ◽  
Michel Gradeck ◽  
Xiaoyang Gaus-Liu ◽  
Thomas Cron ◽  
Beatrix Fluhrer ◽  
...  

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.


2019 ◽  
Vol 2019 ◽  
pp. 1-10 ◽  
Author(s):  
Weitong Li ◽  
Lei Yu ◽  
Jianli Hao ◽  
Mingrui Li

Passive safety system is the core feature of advanced nuclear power plant (NPP). It is a research hotspot to fulfill the function of passive safety system by improving the NPP natural circulation capacity. Considering that the flow behaviors of stopped pump pose a significant effect on natural circulation, both experimental and computational fluid dynamics (CFD) methods were performed to investigate the flow behaviors of a NPP centrifugal pump under natural circulation condition with a low flow rate. Since the pump structure may lead to different flows depending on the flow direction, an experimental loop was set up to measure the pressure drop and loss coefficient of the stopped pump for different flow directions. The experimental results show that the pressure drop of reverse direction is significantly greater than that of forward direction in same Reynolds number. In addition, the loss coefficient changes slightly while the Reynolds number is greater than 8 × 104; however, the coefficients show rapid increase with the decrease in Reynolds number under lower Reynolds number condition. According to the experimental data, an empirical correlation of the pump loss coefficient is obtained. A CFD analysis was also performed to simulate the experiment. The simulation provides a good accuracy with the experimental results. Furthermore, the internal flow field distributions are obtained. It is observed that the interface regions of main components in pump contribute to the most pressure losses. Significant differences are also observed in the flow field between forward and reverse condition. It is noted that the local flows vary with different Reynolds numbers. The study shows that the experimental and CFD methods are beneficial to enhance the understanding of pump internal flow behaviors.


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