Development of the Severe Accident Evaluation Method on Second Coolant Leakages From the PHTS in a Loop-Type Sodium-Cooled Fast Reactor

2017 ◽  
Author(s):  
Fumiaki Yamada ◽  
Yuya Imaizumi ◽  
Masahiro Nishimura ◽  
Yoshitaka Fukano ◽  
Mitsuhiro Arikawa
Keyword(s):  
Sodium Level ◽  
Evaluation Method ◽  
Heat Removal ◽  
Transient Behavior ◽  
Severe Accident ◽  
Accident Condition ◽  
Power Operation ◽  
Heat Removal System ◽  
Accident Sequences ◽  
Removal System

The loss-of-reactor-level (LORL) is a typical accident condition causing a severe accident (SA) in sodium-cooled fast reactors (SFRs). In the loop-type SFR Monju, important cause of the LORL is the second coolant leakage at the low elevation of the primary heat transport system (PHTS), which occurs in cold standby in a different loop from that of the first coolant leakage in rated power operation because of excessive declining of the sodium level. This study developed an evaluation method for the LORL with considering countermeasures to prevent LORL: i.e., pumping sodium up into reactor vessel (RV) and interruption of sodium flowing out by siphon effect. To evaluate the effectiveness of the countermeasures, the transient behavior in the RV sodium level was analytically studied in representative accident sequences. The representative sequences with lowest sodium level were selected by considering combinations of possible coolant leakage positions. The evaluation result clarified that the LORL can be prevented by conducting the above-mentioned countermeasures to maintain the RV sodium level sufficient for the operation of decay heat removal system, even after the second coolant leakage of PHTS.

Fuzzy PSA evaluation method for passive residual heat removal system

2012 ◽  
Vol 247 ◽  
pp. 230-235 ◽  
Author(s):  
Tao Zhou ◽  
Juan Chen ◽  
Feng Luo ◽  
Wanxu Cheng
Keyword(s):  
Evaluation Method ◽  
Heat Removal ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

Study on Hydrogen Risk of Spent Fuel Compartment Induced by Containment Venting

2017 ◽  
Author(s):  
Yabing Li ◽  
Xuewu Cao
Keyword(s):  
Control System ◽  
Heat Removal ◽  
Compartment Model ◽  
Severe Accident ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Spent Fuel Pool ◽  
Heat Removal System ◽  
And Control ◽  
Removal System

Hydrogen risk in the spent fuel compartment becomes a matter of concern after the Fukushima accident. However, researches are mainly focused on the hydrogen generated by spent fuels due to lack of cooling. As a severe accident management strategy, one of the containment venting paths is to vent the containment through the normal residual heat removal system (RNS) to the spent fuel compartment, which will cause hydrogen build up in it. Therefore, the hydrogen risk induced by containment venting for the spent fuel compartment is studied for advanced passive PWR in this paper. The spent fuel pool compartment model is built and analyzed with integral accident analysis code couple with the containment analysis. Hydrogen risk in the spent fuel pool compartment is evaluated combining with containment venting. Since the containment venting is mainly implemented in two different strategies, containment depressurization and control hydrogen flammability, these two strategies are analyzed in this paper to evaluated the hydrogen risk in the spent fuel compartment. Result shows that there will not be significate hydrogen built up with the hydrogen control system available in the containment. However, if the hydrogen control system is not available, venting into the spent fuel pool compartment will cause a certain level of hydrogen risk there. Besides, suggestions are made for containment venting strategy considering hydrogen risk in spent fuel pool compartment.

Experimental and Computational Studies of LNPP-2 Passive System for Severe Accident Management

2010 ◽  
Author(s):  
Valery G. Sidorov ◽  
Vladimir Bezlepkin ◽  
Sergej Alekseev ◽  
Sergey Semashko ◽  
Igor Ivkov ◽  
...  
Keyword(s):  
Heat Removal ◽  
Severe Accident ◽  
Primary Circuit ◽  
Steam Generators ◽  
Final Heat ◽  
Spray System ◽  
Heat Absorber ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

The project of nuclear station LNPP-2 with a reactor power plant VVER type by electrical power 1200 MVt involves a number of new design solutions to increase of parameters of safety. The passive containment heat removal system and heat removal system via steam generators is including of number of such solutions. Passive heat removal system via steam generators (PHRS/SG) is assigned for remove of residual heat of reactors core to final heat absorber (atmosphere) through a secondary circuit at DEC accident. The system PHRS/SG duplicates cooling-down system via SG to final heat absorber in case of impossibility of realization of its design functions. Containment heat removal system (PHRS/C) is assigned for remove of residual heat from containment in accidents with heat-transfer emissions from primary circuit. PHRS/C duplicates functions of a spray system to reduce of pressure under containment in case of spray system failure. In the substantiation of passive security systems the complex in SPbAEP of computational and experimental analysis was executed, the main results of which are shown in the present report.

scholarly journals Analysis on passive residual heat removal system with heat pipes for longterm decay heat removal of small lead-based reactor

2021 ◽  
Vol 236 ◽  
pp. 01018
Author(s):  
Chongju Hu ◽  
Wangli Huang ◽  
Zhizhong Jiang ◽  
Qunying Huang ◽  
Yunqing Bai ◽  
...  
Keyword(s):  
Three Dimensional ◽  
Heat Removal ◽  
Heat Pipes ◽  
Decay Heat ◽  
The Core ◽  
Accident Condition ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

.A lead-based reactor with employing heat pipes as passive residual heat removal system (PRHRS) for longterm decay heat removal was designed. Three-dimensional computational fluid dynamics (CFD) software FLUENT was adopted to simulate the thermal-hydraulic characteristics of the PRHRS under Station-Black-Out (SBO) accident condition. The results showed that heat in the core could be removed smoothly by the PRHRS, and the core temperature difference is less than 20 K.

Approach for the Analysis of a Shutdown Scenario in the Framework of the Validation of Severe Accident Management Guidelines in Belgium

2009 ◽  
Author(s):  
Antonio Cipollaro ◽  
Laurent Sallus
Keyword(s):  
Heat Removal ◽  
The United States ◽  
Severe Accident ◽  
Management Guidelines ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Full Power ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

During last four years, in the framework of the periodic safety review of the Belgian Nuclear Power Plants, the Severe Accident Management Guidelines implemented in Belgium have been involved in a series of detailed validation exercises as suggested by the Westinghouse Owner Group SAMG Scenario Templates. The purpose of this task is essentially to evaluate the severe accident management capabilities of the units and to ensure that personnel in the utility’s emergency response organization (crisis team and eventually the control room staff for certain type of accidents) are trained with the use of the above mentioned guidelines. The supporting calculations to the validation exercises have been performed by Tractebel Engineering by means of the MELCOR 1.8.5 code, which is developed under the sponsorship of the United States Nuclear Regulatory Commission (USNRC). Most of the implemented scenarios and related validation exercises account for full power operating states and are based on previous PSA studies. These included Station Black-Out accidents (SBO), Small Break Loss of Coolant Accidents (SBLOCA), Large Break Loss of Coolant Accidents (LBLOCA), and Interface System Loss of Coolant Accidents (ISLOCA), possibly including additional losses of available emergency safeguards features (ECCS, containment sprays, fan coolers, chemical and volumetric control system). In order to cover the entire spectrum of possible scenarios, it has been judged necessary to consider also a type of accident not originated at nominal power but initiated while the plant is in shutdown conditions. The specific Plant Operating State characterizing this scenario has been defined by a mid-loop operation with the reactor pressure vessel head still in place, and including the opening of the pressurizer manhole, the installation of the nozzle dams in all steam generators, the isolation of the reactor building, and the operation of the Residual Heat Removal system. The initiating event of this accident is the loss of the Residual Heat Removal system one day after the normal reactor shutdown. A point demanding a special attention is the fact the entry criterion to redirect towards the opening of the SAMG (based on core exit temperature measurement in full power states) does not straightforward apply in this case and an alternative criterion is necessary. In particular this paper presents the approach and results obtained accounting for the proposed criterion based on the launch of the internal emergency plan and on the timing for the crisis team to be operational and take the decision.

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