Evaluation of an accident management strategy using an emergency water injection in a reference PWR SFP

Ahn, Kwang-Il,Jung, YongHun,Shin, Jae-Uk,Kim, Won-Tae Elsevier 2018 Annals of nuclear energy Vol.113 No.-

<P><B>Abstract</B></P> <P>The Fukushima accident on March 11, 2011 has shown the relevance of examinations of severe accident inside a spent fuel pool (SFP) during beyond-design-basis external events, and the necessity for provisions to cope effectively with such events through a relevant severe accident management (SAM) strategy. Although the low decay heat of fuel assemblies and the considerable water inventory in an SFP can slow the progress of an accident compared to an accident in the reactor core, the numerous number of fuel assemblies stored inside it and the fact that the SFP building is not leak-tight present the potential for the formation of a direct path for fission products to rise from the SFP into the environment (i.e., a much greater severe accident risk). The purpose of this paper is to assess the effectiveness and success conditions of an emergency makeup water injection strategy, which is being as a representative SFP SAM measure after the Fukushima accident. Two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) were considered in the analysis. For the foregoing SAM strategies, the analysis results and relevant insights are summarized in relation to two major aspects: (a) the key events of the progression of an accident (such as the exposure, heat-up and degradation of the fuel assemblies; the generation of combustible gases such as Hydrogen; and the over-pressurization of the SFP building) and (b) the release of radiological fission products (such as Cesium and Iodine) into the environment. A simulation tool for severe accidents, MELCOR1.8.6, was used in the present analysis.</P> <P><B>Highlights</B></P> <P> <UL> <LI> A SAM strategy for SFP using a water injection is assessed for a PWR. </LI> <LI> Loss-of-cooling/inventory accidents and different reactor operating modes are considered in the analysis. </LI> <LI> Analysis results are given for severe accident progressions and the release of fission products. </LI> <LI> Key insights obtained through the analysis are summarized from a SFP SAM position. </LI> </UL> </P>

CSPACE for a simulation of core damage progression during severe accidents

송진호,손동건,배준호,배성원,하광순,정법동,최유정 한국원자력학회 2021 Nuclear Engineering and Technology Vol.53 No.12

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling ofverified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclearpower plants) and core damage progression code of COMPASS (Core Meltdown Progression AccidentSimulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes,while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels andreactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, coriumbehavior in the lower head are added to COMPASS. Then, an interface module for the data transferbetween two codes was developed to enable coupling. An implicit coupling scheme of wall heat transferwas applied to prevent fluid temperature oscillation. To validate the performance of newly developedcode CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, coredamage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure werereasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction ofsevere accident progression by detailed review of analysis results and a qualitative comparison with theresults of previous MELCOR analysis.

중수로 원전의 발전소정전사고시 격납건물여과배기설비를 고려한 열수력거동 분석

이성한(Sung-Han Lee),김진혁(Jin-Hyuck Kim) 대한기계학회 2015 대한기계학회 춘추학술대회 Vol.열공학 No.-

After the Fukushima accidents, it has been one of the key issues to keep the integrity of the containment under a severe accident. In 2012, containment filtered venting system (CFVS) first installed at Wolsong unit 1 in Korea can be used to depressurizethe containment during a severe accident such as Station Blackout (SBO). In addition, the key function of CFVS is to reduce the radioactive material releasing from the containment to the environment through a high-efficiency scrubber and filters. CFVS is operated when the containment pressure exceeds the design pressure (124 kPa(g)) of the containment building, the operator opens CFVS isolation valves. Then steam and air in the containment flows out to the environment through the CFVS and the containment pressure decreases below 50 kPa(g). The SBO accident is chosen to analyze the depressurization performance of Wolsong unit 1 in consideration of the CFVS operation. The thermal-hydraulic behavior in containment of Wolsong unit 1 was evaluated using the MELCOR code developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC). The geometry of Wolsong unit 1 was modeled with one cell divided into 4 control volumes such as containment, CFVS, CFVS building and environment, and the simulations were run up to 260,000 sec (72 hrs). Also, in order to carry out a sensitivity study, the flow path area as a variable parameter is chosen as 0.0065 ㎡, where this area indicates a venting size of the CFVS for 8 inch, 16 inch, 16 inch and 32 inch, respectively. The results show that the containment pressure is repeatedly rising and falling but it decreases gradually after 46 hrs in case of operating CFVS. On the other hand, the pressure continuously increases and then exceeds the containment failure pressure (427 kPa(a)) at 45.7 hrs without operating CFVS. That is, the containment pressure is considerably decreased and the integrity of the containment could be maintained when CFVS was operated. Therefore, it seems that CFVS has the capacity to keep the containment pressure below the design pressure during SBO. In addition, there are large differences in the containment pressure depending on venting size of CFVS. We found that the decreasing rate of the pressure in the containment and water level in CFVS depends on the venting size of CFVS. In the future, analyses of aerosols, fission product, and radioactive material behavior in containment and remove performance of radionuclide in CFVS are planning to be conducted.

중수로 원전의 격납건물여과배기계통 성능평가

이성한(Sung-Han Lee) 대한기계학회 2015 대한기계학회 춘추학술대회 Vol.2015 No.11

After the Fukushima accidents, it has been one of the key issues to keep the integrity of the containment during a severe accident. In Korea, adding containment filtered venting system (CFVS) installed as one of the Fukushima actions to an existing nuclear power plant has been suggested as one approach to mitigate the effects of a severe accident such as Loss of Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR) and Station Blackout (SBO), etc. CFVS was first installed at Wolsong unit 1 which is the CANDU type reactor. CFVS which does not require an external source of power is used to prevent loss of containment integrity as a result of over-pressurization and relieve containment pressure for the conditions that could be present in a severe accident, and provide reliable venting function for containment pressure control. CFVS is designed to open and to close isolation valves passively by an operator. CFVS is operated when the containment pressure exceeds the design pressure (225 ㎪(a)) and is closed when the containment pressure decreases below 150 ㎪(a) [1]. In addition, the key function of CFVS is to reduce the radioactive material releasing from the containment to the environment through a high-efficiency scrubber and filters. That enables pressure to be reduced using a filtered system that retains and recirculates airborne radioactivity within containment and operates passively without the need for a power supply. Compared with the pressurized water reactor (PWR) containment, the containment of CANDU reactor has lower resistance to internal pressure increase because they have no steel liner inside and design pressure itself is lower. In this regard, the analysis of the thermal-hydraulics behavior of CANDU reactor during a severe accident is needed to evaluate the containment integrity. Therefore, the aim of this study is to evaluate the depressurization performance of Wolsong unit1 considering CFVS operation during SBO using the MELCOR code developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC). In addition, in order to evaluate the effects of the CFVS performance, a sensitivity study depending on the different venting area of the CFVS was conducted. Finally, an analysis of the CFVS performance to evaluate the effects of filtering and scrubbing of radioactive material during a severe accident is important. The effects of filtration efficiencies for the filters of CFVS is evaluated and then a sensitivity study depending on decontamination factor (DF) of the filters was carried out.

Performance evaluation of Accident Tolerant Fuel under station blackout accident in PWR nuclear power plant by improved ISAA code

Zhang Bin,Gao Pengcheng,Xu Tao,Gui Miao,Shan Jianqiang 한국원자력학회 2022 Nuclear Engineering and Technology Vol.54 No.7

The Accident Tolerant Fuel (ATF) is a new concept of fuel, which can not only withstand the consequences of the accident for a longer time, but also maintain or improve the performance under operating conditions. ISAA is a self-developed severe accident analysis code, which uses modular structures to simulate the development processes of severe accidents in nuclear plants. The basic version of ISAA is developed based on UO2eZr fuel. To study the potential safety gain of ATF cladding, an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding. The results show that ATF cladding enable the core to maintain a longer time compared to zirconium alloy cladding, thereby enhancing the accident mitigation capability. Meanwhile, the generation of hydrogen is significantly reduced and delayed, which proves that ATF can improve the safety characteristics of the nuclear reactor.

월성 1 호기 발전소정전사고시 격납건물여과배기설비 성능 평가

이성한(Sung-Han Lee),김진혁(Jin-Hyuck Kim) 대한기계학회 2014 대한기계학회 춘추학술대회 Vol.2014 No.11