scholarly journals Extended Station Blackout Coping Capabilities of APR1400

2014 ◽  
Vol 2014 ◽  
pp. 1-10 ◽  
Author(s):  
Sang-Won Lee ◽  
Tae Hyub Hong ◽  
Mi-Ro Seo ◽  
Young-Seung Lee ◽  
Hyeong-Taek Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Water Injection ◽  
Severe Accident ◽  
Injection Strategy ◽  
Safety Improvement ◽  
External Injection ◽  
Station Blackout ◽  
The Government ◽  
External Water

The Fukushima Dai-ichi nuclear power plant accident shows that an extreme natural disaster can prevent the proper restoration of electric power for several days, so-called extended SBO. In Korea, the government and industry performed comprehensive special safety inspections on all domestic nuclear power plants against beyond design bases external events. One of the safety improvement action items related to the extended SBO is installation of external water injection provision and equipment to RCS and SG. In this paper, the extended SBO coping capability of APR1400 is examined using MAAP4 to assess the effectiveness of the external water injection strategy. Results show that an external injection into SG is applicable to mitigate an extended SBO scenario. However, an external injection into RCS is only effective when RCS depressurization capacity is sufficiently provided in case of high pressure scenarios. Based on the above results, the technical basis of external injection strategy will be reflected on development of revised severe accident management guideline.

Severe Accident Context Quantification for Long-Term Station Blackout in Boiling Water Reactor Nuclear Power Plants

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Gueorgui I. Petkov ◽  
Monica Vela-Garcia
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Accident Scenario ◽  
Station Blackout ◽  
Fukushima Daiichi

The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the performance evaluation of teamwork (PET) procedure for dynamic context quantification and determination of alternatives, coordination, and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions with the use of thermo-hydraulic (TH) model and severe accident (SA) codes (melcor and maap). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and a hypothetic unmitigated LT SBO at peach bottom #1 boiling water reactor (BWR) reactor nuclear power plants (NPPs). The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, “State-of-the-Art Reactor Consequence Analysis” and TH calculations made by using maap code at the EC Joint Research Centre.

scholarly journals A pre-test analysis of ATLAS SBO with RCP seal leakage using MARS code

2021 ◽  
Vol 7 (4) ◽  
pp. 26-33
Author(s):  
Quang Huy Pham ◽  
Sang Yong Lee ◽  
Seung Jong Oh
Keyword(s):  
Coping Strategies ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cooling System ◽  
Water Injection ◽  
Cooling Water ◽  
Feed Water ◽  
Test Loop ◽  
Station Blackout

The accident in Fukushima Daiichi nuclear power plants shows the important of developing coping strategies for extended station blackout (SBO) scenarios of the nuclear power plants (NPPs). Many NPPs in United State of America are applying FLEX approach as main coping strategies for extended station blackout (SBO) scenarios. In FLEX strategies, outside water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. This study presents a pretest calculation using MARS code for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. In the calculation, the turbinedriven auxiliary feed water pumps (TDAFPs) are firstly used after SBO initiation. Then, the outside cooling water injection method is used for long term cooling. In order to minimize operator actions and satisfy requirements of APR1400 emergency operation procedure (EOP), the SGs Atmospheric Dump Valve (ADV) opening ratio, auxiliary feed water (AFW) and outside cooling water injection flow rates were investigated to have suitable values. The analysis results would be useful for performing the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection.

Development of a Water Injection System Fusible Plug for Severe Accidents at Nuclear Power Plants

2020 ◽  
Author(s):  
Sei Hirano ◽  
Daisuke Hirasawa ◽  
Yoshihisa Kiyotoki ◽  
Keisuke Sakemura ◽  
Keiji Sasaki ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Reliability ◽  
Water Injection ◽  
Severe Accident ◽  
Injection System ◽  
Normal Operation ◽  
Autonomous Operation ◽  
Performance Requirements

Abstract Background: When terminal stage of Severe Accident (SA) with no coolant injection at a nuclear power plant, the equipment that has cooled and solidified through water injection to a molten core that has ex-vessel and fallen outside of the pressure vessel will then be required to operate autonomously by heat detection, without external signals or power (e.g. electricity, air). The fusible plug operation is triggered by fusible alloy which receives heat from molten core and will melt. Because the fusible plug is also the boundary of Suppression Pool (S/P), high reliability is required for sealing performance. It is for that reason that Hitachi GE Nuclear Energy Ltd. (Hitachi-GE) has developed a fusible plug to serve as a device necessary to operate this system. Features of the Fusible Plug: The autonomous operation of the fusible plug is triggered by the melting of a fusible alloy, which is part of the fusible plug. However, the fusible alloy has a remarkably low mechanical strength and therefore is not suitable as a strength member. As such, it is necessary to ensure reliable plug sealing without applying a load to the fusible alloy so as to prevent the fusible plug from malfunctioning during normal operation. Therefore, to reduce the load to be applied to the fusible alloy, Hitachi-GE has developed a fusible plug structure that operates autonomously by detecting the ambient temperature without using the fusible alloy as a strength member. We have performed a verification test using this fusible plug and confirmed that it satisfies the predetermined performance requirements. Future Actions: Hitachi-GE is holding discussions on using the fusible plug at nuclear power plants in Japan. In the future, we plan to expand to the overseas.

Effectiveness of RCS Depressurization Strategy for Severe Accident Management

2009 ◽  
Author(s):  
Changwook Huh ◽  
Namduk Suh ◽  
Goon-Cherl Park
Keyword(s):  
Survival Time ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Time Interval ◽  
Counter Measures ◽  
Accident Management ◽  
Station Blackout ◽  
Safety Features ◽  
Reactor Pressure

In developing the severe accident management guideline (SAMG), it was highly considered to maintain the integrity of reactor pressure vessel (RPV) as a key strategy aimed to reduce the risk of containment failure and fission product releases into the environment effectively. For the operating nuclear power plants with no dedicated safety features for the severe accident management (SAM), the improvement of the current mitigative strategy in SAMG domain could be one of counter-measures to extend the survival time of RPV. In this study, the effectiveness of the RCS depressurization to delay the RPV survival time with different RCS depressurization rate was evaluated for station blackout (SBO) accident assuming only SIT is available for Uljin unit 1 plant by MELCOR 1.8.5 code. According to the analysis results, it was shown that the conditions for RCS depressurization such as depressurization capacity and the time interval are the key elements to extend the RPV integrity in such a way to earn time for restoring the heat sink in order to prevent the accident propagation to the RPV failure.

The Analysis of Effectiveness of BWR/6 URG Strategies

2017 ◽  
Vol 865 ◽  
pp. 701-706
Author(s):  
Ting Yi Wang ◽  
Yu Ting Hsu ◽  
Shao Wen Chen ◽  
Jong Rong Wang ◽  
Chun Kuan Shih ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Injection ◽  
Water Source ◽  
Injection Rate ◽  
Power Company ◽  
Pressure Water ◽  
Station Blackout

After the Fukushima Daichii accidents, Taiwan Power Company developed a strategy to cope with such extended SBO (Station Blackout) cases at nuclear power plants, which called URG. The MAAP and PCTRAN were used to perform the study for Kuosheng BWR/6 nuclear power plant (NPP) ultimate response guideline (URG). The main actions of URG are the depressurization and low pressure water injection of the reactor and the venting of the containment. This study focuses to confirm the URG efficiency. The analysis results depict that following the URG, the fuel can be covered by the coolant, no exposure. It can also prevent the radiation release and the large evacuation. It indicated that Kuosheng NPP was at the safe situation. It shows that the two-step depressurizations can extend the time of the preparation of alternate water source. The minimum injection rate to prevent the fuel to expose is 192 gpm in MAAP.

Education and Exercise for Severe Accident Management at KANSAI

2017 ◽  
Author(s):  
Takumi Kawahara ◽  
Tsugio Shiozawa ◽  
Tsutomu Nishioka ◽  
Yukimoto Shimominami ◽  
Hiroaki Onooka
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Exercise Program ◽  
Severe Accident ◽  
Staff Members ◽  
Power Company ◽  
Safety Improvement ◽  
Management Ability ◽  
Accident Management ◽  
Response Organization

In case of a severe accident, all the staff members of a nuclear power plant (NPP), members of the Emergency Response Organization (ERO) on site as well as operators in the main control room (MCR) are required to take necessary actions to mitigate the consequential effects of the accident. Therefore, Nuclear Engineering Ltd (NEL) has been implementing education and exercise for severe accident (SA) management both at nuclear power plants and Nuclear Power Division of Kansai Electric Power Company (KANSAI) since FY 2014. For the education of commanders who take a lead at the ERO in case of an accident, table top exercise is provided by using simulators developed by NEL, including functions to respond to a SA involving core melt. Continuous implementation of this education and exercise program is expected to enhance KANSAI’s severe accident management ability and their voluntary safety improvement activities in the future.

scholarly journals Simulation of a Station Blackout Accident for the SMART Using the CINEMA Code

2020 ◽  
Vol 8 ◽  
Author(s):  
Hyoung Tae Kim ◽  
Jin Ho Song ◽  
Rae-Joon Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Hydrogen Generation ◽  
Alternate Current ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Active Safety Systems ◽  
Station Blackout ◽  
Safety Systems

SMART is a small-sized integral type PWR containing major components within a single reactor pressure vessel. Advanced design features implemented into SMART have been proven or qualified through experience, testing, or analysis according to the applicable approved standards. After Fukushima accident, a rising attention is posed on the strategy to cope with a Station Blackout (SBO) accident, which is one of the representative severe accidents related to the nuclear power plants. The SBO is initiated by a loss of all offsite power with a concurrent failure of both emergency diesel generators. With no alternate current power source, most of the active safety systems that perform safety functions are not available. The purpose of SBO analysis in this paper is to show that the integrity of the containment can be maintained during a SBO accident in the SMART (System-integrated Modular Advanced ReacTor). Therefore, the accident sequence during a SBO accident was simulated using the CINEMA-SMART (Code for INtegrated severe accidEnt Management and Analysis-SMART) code to evaluate the transient scenario inside the reactor vessel after an initiating event, core heating and melting by core uncovery, relocation of debris, reactor vessel failure, discharge of molten core, and pressurization of the containment. It is shown that the integrity of the containment can be maintained during a SBO accident in the SMART reactor. It has to be mentioned that the assumptions used in this analysis are extremely conservative that the passive safety systems of PSIS and PRHRS were not credited. In addition, as ANS73 decay heat with 1.2 multiplier was used in this analysis, actual progression of the accident would be much slow and amount of hydrogen generation will be much less.

scholarly journals Application of MELCOR to Westinghouse 4-loop PWR severe accident and evaluation of RPV lower head performance

2014 ◽  
Vol 4 (3) ◽  
pp. 1-6
Author(s):  
Dai Dien Le ◽  
Thi Hoa Bui ◽  
Thi Huong Vo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Computer Code ◽  
Severe Accident ◽  
Ac Power ◽  
Lower Head ◽  
Active Safety Systems ◽  
Station Blackout ◽  
Control Volumes ◽  
Safety Systems

In this study, MELCOR computer code is used to simulate the progression of a severe accident initiated from station blackout (SBO) accident for a Westinghouse 4-loop PWR. The hydraulic system is modeled using control volumes and flow paths. The reactor pressure vessel and internals, the primary loops with a pressurizer, steam generators, containment and accumulators are simulated for steady state in a good agreement with reference data. The two scenarios concerning SBO are investigated. The first scenario simulates RCP seal leakage during SBO and the other is SBLOCA to highlight an effectiveness of accumulators as well as to compare with the first simulation. All active safety systems which depend on AC power are assumed to be unavailable in this analysis. The main result of the study is an evaluation of RPV lower head integrity during severe accidents. This is preliminary work and expected to give the experience for further studies in the severe accident in nuclear power plants.

scholarly journals Containment Depressurization Capabilities of Filtered Venting System in 1000 MWe PWR with Large Dry Containment

2014 ◽  
Vol 2014 ◽  
pp. 1-10 ◽  
Author(s):  
Sang-Won Lee ◽  
Tae-Hyub Hong ◽  
Yu-Jung Choi ◽  
Mi-Ro Seo ◽  
Hyeong-Taek Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Noble Gas ◽  
Decontamination Factor ◽  
Severe Accident ◽  
Spray System ◽  
Discharge Flow Rate ◽  
Safety Improvement ◽  
Accident Conditions ◽  
Containment Integrity

After the Fukushima Daiichi nuclear power plant accident, the Korean government and nuclear industries performed comprehensive safety inspections on all domestic nuclear power plants against beyond design bases events. As a result, a total of 50 recommendations were defined as safety improvement action items. One of them is installation of a containment filtered venting system (CFVS) or portable backup containment spray system. In this paper, the applicability of CFVS is examined for OPR1000, a 1000 MWe PWR with large dry containment in Korea. Thermohydraulic analysis results show that a filtered discharge flow rate of 15 [kg/s] at 0.9 [MPa] is sufficient to depressurize the containment against representative containment overpressurization scenarios. Radiological release to the environment is reduced to10-3considering the decontamination factor. Also, this cyclic venting strategy reduces noble gas release by 50% for 7 days. The probability of maintaining the containment integrity in level 2 probabilistic safety assessment (PSA) initiating events is improved twofold, from 43% to 87%. So, the CFVS can further improve the containment integrity in severe accident conditions.

scholarly journals Application of MELCOR to Westinghouse 4-loop PWR severe accident and evaluation of RPV lower head performance

2014 ◽  
Vol 4 (3) ◽  
pp. 19-28
Author(s):  
Dai Dien Le ◽  
Thi Hoa Bui ◽  
Thi Huong Vo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Computer Code ◽  
Severe Accident ◽  
Ac Power ◽  
Lower Head ◽  
Active Safety Systems ◽  
Station Blackout ◽  
Control Volumes ◽  
Safety Systems

In this study, MELCOR computer code is used to simulate the progression of a severe accident initiated from station blackout (SBO) accident for a Westinghouse 4-loop PWR. The hydraulic system is modeled using control volumes and flow paths. The reactor pressure vessel and internals, the primary loops with a pressurizer, steam generators, containment and accumulators are simulated for steady state in a good agreement with reference data. The two scenarios concerning SBO are investigated. The first scenario simulates RCP seal leakage during SBO and the other is SBLOCA to highlight an effectiveness of accumulators as well as to compare with the first simulation. All active safety systems which depend on AC power are assumed to be unavailable in this analysis. The main result of the study is an evaluation of RPV lower head integrity during severe accidents. This is preliminary work and expected to give the experience for further studies in the severe accident in nuclear power plants.

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