Safety assessment of severe accident mitigating system “Passive Residual Heat Removal System of secondary side” in K-2/K-3 NPP

Author(s):  
Muhammad Abu Bakar ◽  
Muhammad Sohail Sarwar
Author(s):  
Valery G. Sidorov ◽  
Vladimir Bezlepkin ◽  
Sergej Alekseev ◽  
Sergey Semashko ◽  
Igor Ivkov ◽  
...  

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.


Author(s):  
Ran Xu

Passive systems are widely used in nuclear power plants, to enhance the inherent safety of the plant. Especially in the accident of station blackout, Passive Residual Heat Removal system (PRHR) which is installed on the secondary side of steam generator (SG), highly resolved the problem of the residual heat removal after reactor trip. The plant safety is improved largely by PRHR, this is the trend of the new plant design. A water tank is commonly mounted on the passive residual heat removal system (PRHR) of steam generator secondary side, which heavily impacts the performance and the startup procedure of PRHR. Moreover, there are many different types of tank layout in various plants design. Therefore it is necessary to study the thermal hydraulic feature of the different type the PRHR tank set. In this paper, the station blackout accident is chosen to verify the performance of the tank set. Then the assessment criteria of PRHR is established from a thermal hydraulic perspective. RELAP5/mod3.2 is adopted to simulate the accident. Based on the performance comparison of different tank sets, the best tank layout is selected, which is selected for the engineering practice. At last, the comparison of the predicted results and the test data is done to verify the performance of selected tank set design.


Author(s):  
Antonio Cipollaro ◽  
Laurent Sallus

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.


2016 ◽  
Vol 89 ◽  
pp. 56-62 ◽  
Author(s):  
Yeon-Sik Kim ◽  
Sung-Won Bae ◽  
Seok Cho ◽  
Kyoung-Ho Kang ◽  
Hyun-Sik Park

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