scholarly journals Assessment of Severe Accident Depressurization Valve Activation Strategy for Chinese Improved 1000 MWe PWR

2013 ◽  
Vol 2013 ◽  
pp. 1-8
Author(s):  
Ge Shao ◽  
Lili Tong ◽  
Xuewu Cao
Keyword(s):  
Heat Transfer ◽  
Natural Circulation ◽  
Cooling Water ◽  
Severe Accident ◽  
Capacity Analysis ◽  
The Core ◽  
Core Cooling ◽  
Accident Sequences ◽  
Delay Progression ◽  
Valve Area

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.

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.

Investigation of Countermeasure Against Local Temperature Rise in Vessel Cooling System in Loss of Core Cooling Test Without Nuclear Heating

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Masato Ono ◽  
Atsushi Shimizu ◽  
Makoto Kondo ◽  
Yosuke Shimazaki ◽  
Masanori Shinohara ◽  
...  
Keyword(s):  
High Temperature ◽  
Cooling System ◽  
Natural Circulation ◽  
Reactor Core ◽  
Severe Accident ◽  
Local Temperature ◽  
Water Cooling ◽  
The Core ◽  
Cooling Tube ◽  
Core Cooling

In the loss of core cooling test using High Temperature engineering Test Reactor (HTTR), the forced cooling of the reactor core is stopped without inserting control rods into the core and, furthermore, without cooling by the vessel cooling system (VCS) to verify safety evaluation codes to investigate the inherent safety of high-temperature gas-cooled reactor (HTGR) be secured by natural phenomena to make it possible to design a severe accident-free reactor. The VCS passively removes the retained residual heat and the decay heat from the core via the reactor pressure vessel (RPV) by natural convection and thermal radiation. In the test, the local temperature was supposed to exceed the limit from the viewpoint of long-term use at the uncovered water-cooling tube without thermal reflectors in the VCS, although the safety of reactor is kept. Through a cold test, which was carried out by non-nuclear heat input from helium gas circulators (HGCs) by stopping water flow in the VCS, the local higher temperature position was specified in the uncovered water-cooling tube of the VCS, although the temperature was sufficiently lower than the maximum allowable working temperature, and the natural circulation of water had an insufficient cooling effect on the temperature of the water-cooling tube below 1°C. Then, a new safe and secured procedure for the loss of core cooling test was established, which will be carried out soon after the restart of HTTR.

Ex-Vessel Coolability Analysis for a Belgian NPP

2009 ◽  
Author(s):  
Jarne R. Verpoorten ◽  
Miche`le Auglaire ◽  
Frank Bertels
Keyword(s):  
State Of The Art ◽  
The State ◽  
Severe Accident ◽  
Research Projects ◽  
Specific Data ◽  
The Core ◽  
Core Damage ◽  
Accident Management ◽  
Lower Head ◽  
Core Cooling

During a hypothetical Severe Accident (SA), core damage is to be expected due to insufficient core cooling. If the lack of core cooling persists, the degradation of the core can continue and could lead to the presence of corium in the lower plenum. There, the thermo-mechanical attack of the lower head by the corium could eventually lead to vessel failure and corium release to the reactor cavity pit. In this paper, it is described how the international state-of-the-art knowledge has been applied in combination with plant-specific data in order to obtain a custom Severe Accident Management (SAM) approach and hardware adaptations for existing NPPs. Also the interest of Tractebel Engineering in future SA research projects related to this topic will be addressed from the viewpoint of keeping the analysis up-to-date with the state-of-the art knowledge.

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.

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 Modeling Strategies to Compute Natural Circulation Using CFD in a VHTR After a LOFA

2012 ◽  
Author(s):  
Yu-Hsin Tung ◽  
Richard W. Johnson ◽  
Ching-Chang Chieng ◽  
Yuh-Ming Ferng
Keyword(s):  
Heat Transfer ◽  
Natural Circulation ◽  
Department Of Energy ◽  
Full Potential ◽  
The Core ◽  
Triso Fuel ◽  
Flow And Heat Transfer ◽  
Local Symmetries ◽  
Fuel Particles ◽  
Global And Local

A prismatic gas-cooled very high temperature reactor (VHTR) is being developed under the next generation nuclear plant program (NGNP) of the U.S. Department of Energy, Office of Nuclear Energy. In the design of the prismatic VHTR, hexagonal shaped graphite blocks are drilled to allow insertion of fuel pins, made of compacted tristructural-isotropic (TRISO) fuel particles, and coolant channels for the helium coolant. One of the concerns for the reactor design is the effects of a loss of flow accident (LOFA) where the coolant circulators are lost for some reason, causing a loss of forced coolant flow through the core. In such an event, it is desired to know what happens to the (reduced) heat still being generated in the core and if it represents a problem for the fuel compacts, the graphite core or the reactor vessel (RV) walls. One of the mechanisms for the transport of heat out of the core is by the natural circulation of the coolant, which is still present. It is desired to know how much heat may be transported by natural circulation through the core and upwards to the top of the upper plenum. It is beyond current capability for a computational fluid dynamics (CFD) analysis to perform a calculation on the whole RV with a sufficiently refined mesh to examine the full potential of natural circulation in the vessel. The present paper reports the investigation of several strategies to model the flow and heat transfer in the RV. It is found that it is necessary to employ representative geometries of the core to estimate the heat transfer. However, by taking advantage of global and local symmetries, a detailed estimate of the strength of the resulting natural circulation and the level of heat transfer to the top of the upper plenum is obtained.

scholarly journals Long-Term Station Blackout Accident Analyses of a PWR with RELAP5/MOD3.3

2013 ◽  
Vol 2013 ◽  
pp. 1-15 ◽  
Author(s):  
Andrej Prošek ◽  
Leon Cizelj
Keyword(s):  
Management Strategies ◽  
Severe Accident ◽  
Stress Tests ◽  
Pressurized Water ◽  
Makeup Water ◽  
Reactor Coolant ◽  
The Core ◽  
Station Blackout ◽  
Accident Sequences

Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pumps (RCPs) leaks assumed) to delay the time before core uncovers and significantly heats up. The most important strategies assumed were primary side depressurization and additional makeup water to reactor coolant system (RCS). For simulations of long term SBO scenarios, including early stages of severe accident sequences, the best estimate RELAP5/MOD3.3 and the verified input model of Krško two-loop PWR were used. The results suggest that for the expected magnitude of RCPs seal leak, the core uncovery during the first seven days could be prevented by using the turbine-driven auxiliary feedwater pump and manually depressurizing the RCS through the secondary side. For larger RCPs seal leaks, in general this is not the case. Nevertheless, the core uncovery can be significantly delayed by increasing RCS depressurization.

Analysis of IP200 Severe Accident Process Response to SBO and Emergency Power Failure

2021 ◽  
Author(s):  
Zhenhang Zheng ◽  
Minjun Peng ◽  
Hao Yu ◽  
Yang Yang
Keyword(s):  
Cooling System ◽  
Cooling Capacity ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Pressurized Water ◽  
Power Failure ◽  
The Core ◽  
Emergency Power ◽  
Core Cooling ◽  
High Level

Abstract Advanced SMRs such as the integrated pressurized water reactor IP200 use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In this thesis, the IP200 severe accident induced by the SBO and emergency power failure was modeled and analyzed using RELAP5 / SCDAP / MOD3.4 code. Based on the steady state calculation, which agrees well with designed values, the SBO accident for transient calculation is carried out. First, the case of the SBO accident without the passive core cooling system was calculated. The progression and scenario in the RPV was simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. Then, mitigation measures PRHRS and CMT were put in at four different time points when the core is began to uncovered, the core is completely uncovered, hydrogen is began to produced, and the molten pool is formed. The results show that putting in mitigation measures before the accident progresses to the point where the core starts to produce hydrogen can ensure that the core does not melt and avoid hydrogen risk.

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