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 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.

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 Моделирование окисления расплава активной зоны ядерного реактора при наличии оксидной корки на поверхности расплава

2021 ◽  
Vol 91 (2) ◽  
pp. 232
Author(s):  
В.Б. Хабенский ◽  
В.И. Альмяшев ◽  
В.С. Грановский ◽  
Е.В. Крушинов ◽  
С.А. Витоль ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Solid Phase ◽  
Reactor Vessel ◽  
Hydrogen Release ◽  
Severe Accident ◽  
Water Steam ◽  
Exothermic Reactions ◽  
Realistic Situation ◽  
Water Reactors

At a severe accident of nuclear power plants with light-water reactors, the most effective way to localize the forming melt (corium) is to keep it in the cooled reactor vessel, the integrity of which depends on the value of heat flux from the melt to the reactor vessel. In this case, one of the critical processes is the melt oxidation by a water steam or a steam-air mixture. It process can lead to a significant increase in the thermal load on the reactor vessel due to a heat of exothermic reactions of oxidation of reducing agents, which presents in the melt, a thickness decreasing of the metallic part of the molten pool, and a hydrogen release. All of these factors strictly depends on the rate of oxidation. When considering the conditions of melt oxidation, it taken into account that for the accepted scenarios of a severe accident, the most realistic situation is the presence of a solid-phase oxide layer (oxidic crust) on the melt surface. Under these conditions, a dependence for calculating the rate of core melt oxidation based on the diffusion model proposed and its validation by using the obtained experimental data performed.

Diagnostics and Prognostics for Safe and Secure Operation of Small Modular Reactors

2011 ◽  
Author(s):  
Pradeep Ramuhalli ◽  
Ryan Meyer ◽  
Leonard Bond
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Economies Of Scale ◽  
Modular Design ◽  
The United States ◽  
Pressure Vessels ◽  
Economy Of Scale ◽  
Small Modular Reactors ◽  
Active Safety Systems ◽  
Safety Systems

Sustainable nuclear power to promote energy security is a key national energy priority. The development of small modular reactors (SMRs) is expected to provide the United States with an economically viable energy option that supports this priority. Small modular reactors (SMR) are typically defined as nuclear reactors that have electrical output less than about 300 MWe [1]. In recent years, SMRs are seeing renewed interest due to several factors: 1. Economy of scale. Modular or grid-appropriate reactors can be used to expand power plants to meet needs [2], resulting in potential economies of scale that larger reactors cannot easily provide. 2. Ease of fabrication/construction. The forging capabilities for the smaller reactor pressure vessels and piping necessary for SMRs are more readily available. This also enables potentially faster and easier construction of SMRs. 3. Usually grid-following. SMRs can ramp up or down production of electricity as needed as the modular design allows for better control of grid-appropriate reactors. 4. Improved safety characteristics. Most current SMR designs rely on passive rather than active safety systems.

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Sumit V. Prasad ◽  
A. K. Nayak
Keyword(s):  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Public Domain ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Critical Issues

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Creation of Automatic System for Monitoring, Forecasting and Control of Condition of Lead-Bismuth (Lead) Coolant and Surfaces of Circuits of Nuclear Power Plants

2009 ◽  
Author(s):  
P. N. Martynov ◽  
R. Sh. Askhadullin ◽  
A. A. Simakov ◽  
A. Yu. Chaban’ ◽  
M. E. Chernov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Heat Removal ◽  
Severe Accident ◽  
Medium Size ◽  
Safety Level ◽  
Nitride Fuel ◽  
Active Safety Systems ◽  
Accident Conditions

Lead-bismuth coolant is preferable for the medium size reactors, since, in contrast to the sodium coolant, it does not interact with water and air, it is radiation resistant, insignificantly activated and it is not combustible [1]. Combination of natural properties of lead-based coolants, mono-nitride fuel, fast reactor neutronics and design approaches used for the reactor core and heat removal system brings SVBR 75/100 NPP [2] to achieve a new safety level and assures its stability without operation of active safety systems even under severe accident conditions. Analysis of possible sequences of the events even under conditions of such severe accidents as addition of total excess reactivity or all pumps trip accompanied by safety system failure leads to the conclusion on that power unit with SVBR 75/100 reactor plant (RP) has high safety level.

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.

scholarly journals Analysis of Reliability-Critical Hydraulic Impact Conditions at WWER-1000 NPP Active Safety Systems

2019 ◽  
pp. 42-45
Author(s):  
V. Skalozubov ◽  
I. Kozlov ◽  
O. Chulkin ◽  
Yu. Komarov ◽  
O. Piontkovskyi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Original Method ◽  
Injection System ◽  
Active Safety ◽  
Pipeline System ◽  
Hydrodynamic Parameters ◽  
Hydraulic Impact ◽  
Active Safety Systems ◽  
Safety Systems

An original method to determine the onset of conditions for reliabilitycritical hydraulic impacts for reliability analysis of active safety systems of nuclear power installations is proposed. The suggested method is based on determining the effect of head-flow characteristic delay onto hydraulic impact preconditions during the changes of pipeline system hydrodynamic parameters under transitional modes (e.g. in pump start-up). The delay time of responses to change in the hydrodynamic system parameters embodies the determining factor of head-flow characteristic’s inertance, depending on both design and technical parameters of system components (including pumps), and the hydrodynamic parameters change rate under transitional modes. Using the proposed method, the analysis of conditions for critical hydraulic impacts is performed for the primary high-pressure safety injection system of serial WWER-1000 nuclear power plants. The analysis results allow a conclusion that for this system the conditions leading to hydraulic impact due to the pump start are not reached. The developed method can be applied to any thermal and nuclear power facilities’ pipeline systems equipped with pumps.

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