scholarly journals Analysis of the Processes in Spent Fuel Pools in Case of Loss of Heat Removal due to Water Leakage

2013 ◽  
Vol 2013 ◽  
pp. 1-11 ◽  
Author(s):  
Algirdas Kaliatka ◽  
Viktor Ognerubov ◽  
Virginijus Vileiniškis ◽  
Eugenijus Ušpuras
Keyword(s):  
Reactor Core ◽  
Heat Removal ◽  
Severe Accident ◽  
Accident Analysis ◽  
Spent Fuel ◽  
Decay Heat ◽  
Management Measures ◽  
Fuel Assemblies ◽  
Accident Management ◽  
Spent Fuel Pool

The safe storage of spent fuel assemblies in the spent fuel pools is very important. These facilities are not covered by leaktight containment; thus, the consequences of overheating and melting of fuel in the spent fuel pools can be very severe. On the other hand, due to low decay heat of fuel assemblies, the processes in pools are slow in comparison with processes in reactor core during LOCA accident. Thus, the accident management measures play a very important role in case of some accidents in spent fuel pools. This paper presents the analysis of possible consequences of fuel overheating due to leakage of water from spent fuel pool. Also, the accident mitigation measure, the late injection of water was evaluated. The analysis was performed for the Ignalina NPP Unit 2 spent fuel pool, using system thermal hydraulic code for severe accident analysis ATHLET-CD. The phenomena, taking place during such accident, are discussed. Also, benchmarking of results of the same accident calculation using ASTEC and RELAP/SCDAPSIM codes is presented here.

Analysis of a Loss of Heat Removal Accident in a PWR Spent Fuel Pool

2014 ◽  
Author(s):  
Xiaoli Wu ◽  
Yapei Zhang ◽  
Wenxi Tian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Flow Rate ◽  
Evaporation Rate ◽  
Heat Removal ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
Reference Case ◽  
Makeup Water ◽  
Fuel Assemblies ◽  
Spent Fuel Pool

The Fukushima Daiichi nuclear accident shows that it is necessary to study potential severe accidents and corresponding mitigation measures for the spent fuel pool (SFP) of a nuclear power plant (NPP). This paper presents a study on the consequences of loss of heat removal accident in the spent fuel pool of a typical pressurized water reactor using the Modular Accident Analysis Program (MAAP5) code. Analysis of uncompensated loss of water due to the loss of heat removal with initial pool water level of 12.2 m (designated as a reference case) has been performed. The analyses cover a broad spectrum of severe accident in the spent fuel pool. Those consequences such as overheating of uncovered fuel assemblies, oxidation of zirconium and hydrogen generation, loss of intactness of fuel rod claddings, and release of radioactive fission product are also analyzed in this paper. Furthermore, as important mitigation measures, the effects of makeup water in SFP on the accident progressions have also been investigated based on the events of spent fuels uncovery. The results showed that spent fuels could be completely submerged and severe accident might be avoided if SFP makeup water system provided water with a mass flow rate higher than evaporation rate defined in the reference case. Although spent fuel assemblies partly exposed due to a mass flow rate of makeup water smaller than the average evaporation rate, continuous steam cooling and radiation heat transfer might maintain the spent fuels coolability as the actual evaporation was balanced by the makeup in a period of time of the order of several days. However, larger makeup rate should be guaranteed to ensure long-term safety of SFP.

Research and Development of Validation and Drill System for Full Scope of Severe Accident Management Guideline

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
Zhifei Yang ◽  
Yali Chen ◽  
Hu Luo
Keyword(s):  
Research And Development ◽  
Computer Code ◽  
Severe Accident ◽  
Accident Analysis ◽  
Spent Fuel ◽  
Accident Management ◽  
Analysis Computer ◽  
Spent Fuel Pool ◽  
Accident Scenarios ◽  
Drill System

To respond to the urgent needs of verification, training, and drill for full scope severe accident management guidelines (FSSAMG) among nuclear regulators, utilities, and research institutes, the FSSAMG verification and drill system is developed. The FSSAMG includes comprehensive scenarios under power condition, shutdown condition, spent fuel pool (SFP) condition, and refueling conditions. This article summarized the research and development of validation and drill system for FSSAMG by using the severe accident analysis computer code modular accident analysis program 5 (MAAP5). Realistic accident scenarios can be verified and exercised in the developed system to support FSSAMG training, drill, examination, and verification.

The Design and CFD Simulation of a New Spent Fuel Pool Passive Cooling System

2013 ◽  
Author(s):  
Cheng Ye ◽  
Minglu Wang ◽  
Mingguang Zheng ◽  
Zhengqin Xiong ◽  
Ronghua Zhang
Keyword(s):  
High Efficiency ◽  
Cfd Simulation ◽  
Cooling System ◽  
Severe Accident ◽  
Passive Cooling ◽  
Spent Fuel ◽  
Decay Heat ◽  
Fuel Assemblies ◽  
Spent Fuel Pool ◽  
Passive Cooling System

Due to the safety issues arising from the Fukushima accident, a novel completely passive spent fuel pool cooling system is proposed using the high-efficiency heat pipe cooling technology that is available in an emergency condition such as a station blackout. This cooling system’s ability to remove the decay heat released by the spent fuel assemblies is evaluated by a computational fluid dynamics (CFD) simulation. The spent fuel pool of CAP1400 (a passive PWR developed in China) is selected as the reference pool, and the passive cooling system is designed for this spent fuel pool. The pool with the passive cooling system is simulated using Fluent 13.0 with 4 million meshes. Four different cases have been studied, and some notable results have been obtained through this work. The simulation results reveal that the passive cooling system effectively removes the decay heat from the SFP with the storage of 15-year-old spent fuel assemblies with emergency reactor core unloading and prevents the burnout of the fuel rods. The results indicate that the water in the SFP will never boil, even in a severe accident with a lack of emergency power and outside aid.

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.

Problems With Cooling of Overheated Fuel Channels in RBMK-Type Reactors

2008 ◽  
Author(s):  
Eugenijus Uspuras ◽  
Algirdas Kaliatka
Keyword(s):  
Heat Removal ◽  
Water Source ◽  
Severe Accident ◽  
Management Guidelines ◽  
Management Measures ◽  
Accident Management ◽  
Feed Strategy ◽  
Significant Probability ◽  
Reactor Types

One of the most dangerous beyond design basis accidents for all types of nuclear reactors is the loss of long-term heat removal from the core. In RBMK-type reactors, this initiating event, which can lead to the worst consequences, has significant probability to occur in comparison to other type of BDBA. The most effective accident mitigation measure in this case is “bleed and feed” strategy — similar as is recommended for other light water reactor types. In this paper the challenges, which are meet in case of cooling of overheated fuel channels in RBMK-type reactors, are discussed. The simulation results of BDBA using RELAP5/MOD3.3 code are presented. Accident management measures (de-pressurization of reactor cooling circuit and injection of water from non-regular water source) are evaluated in respect of dangerous pressure increase and thermal shock in fuel channels. These results were used during development of severe accident management guidelines for RBMK-1500 at Ignalina NPP.

Corium and Debris Coolability Studies Performed in the Severe Accident Research Network of Excellence (SARNET2)

2012 ◽  
Author(s):  
Alexei Miassoedov ◽  
Th. Walter Tromm ◽  
Jonathan Birchley ◽  
Florian Fichot ◽  
Weimin Ma ◽  
...  
Keyword(s):  
Reactor Core ◽  
Safety Evaluation ◽  
Severe Accident ◽  
System Level ◽  
Research Network ◽  
Work Package ◽  
Management Measures ◽  
Accident Management ◽  
Lower Head ◽  
Accident Research

The motivation of the work performed within the work package “Corium and Debris Coolability” of the Severe Accident Research Network of Excellence (SARNET) is to reduce or possibly solve the remaining uncertainties on the efficiency of cooling reactor core structures and materials during severe accidents, either in the core, in the vessel lower head or in the reactor cavity, so as to limit the progression of the accident. This can be achieved either by ensuring corium retention within the reactor pressure vessel or at least by limiting the corium progression and the rate of corium release into the cavity. These issues are to be covered within the scope of accident management for existing reactors and within the scope of design and safety evaluation of future reactors. The specific objectives are to create and enhance the database on debris formation, debris coolability and corium behavior in the lower head, to develop and validate the models and computer codes for simulation of in-vessel debris bed and melt pool behavior, to perform reactor scale analysis for in-vessel corium coolability and to assess the influence of severe accident management measures on in-vessel coolability. The work being performed within this work package comprises experimental and modeling activities with strong cross coupling between the tasks. Substantial knowledge and understanding of governing phenomena concerning coolability of intact rod-like reactor core geometry was obtained in previous projects. Hence the main thrust of experimental and modeling efforts concentrates mainly on the study of formation and cooling of debris beds in order to demonstrate effective cooling modes, cooling rates and coolability limits. Modeling efforts have been aimed at assessing and validating the models in system-level and detailed codes for core degradation, oxidation and debris behavior. The paper describes the work performed up to now and summarizes the main results achieved so far.

Innovative Practice of Severe Accidents Measures in CAP1400

2017 ◽  
Author(s):  
Likai Fang ◽  
Xin Liu ◽  
Guobao Shi
Keyword(s):  
Nuclear Power ◽  
Heat Removal ◽  
Severe Accident ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
Fukushima Accident ◽  
Power Sources ◽  
Severe Accidents ◽  
Accident Management ◽  
Accident Conditions

CAP1400 is GenIII passive PWR, which was developed based on Chinese 40 years of experience in nuclear power R&D, construction&operation, as well as introduction and assimilation of AP1000. Severe accidents prevention and mitigation measures were systematically considered during the design and analysis. In order to accommodate high power and further improve the safety of the plant, also considering feedback from Fukushima accident, some innovative measures and design requirements were also applied. Based on the probabilistic&deterministic analysis and engineering judgment, considerable severe accidents scenarios were considered. Both severe accidents initiated at power and shutdown condition were analyzed. Insights were also obtained to decide the challenge to the plant. All known severe accidents phenomena and their treatment were considered in the design. In vessel retention (IVR) was applied as one of the severe accident mitigation measures. To improve the margin of IVR success and verify the heat removal capability through reactor pressure vessel, both design innovative measures and experiments were used. The melt pool behavior and corium pool configuration were also studied by using CFD code and thermodynamic code. Hydrogen risk was mitigated by installation of hydrogen igniters, which were comprised of two serials, and were powered by multiple power sources. To further improve the safety, six extra hydrogen passive recombiners were also added in the containment. Hydrogen risk was analyzed both inside containment and outside containment considering leakage effect. Other severe accident phenomena were also considered by designed or analyzed to show the containment robustness to accommodate it. As one of the Fukushima accident feedback, full scope severe accident management guideline were developed by considering both power condition and shutdown condition, accident management for spent fuel pool was also considered. As the basis of accident management during severe accidents, survivability of equipments and instruments that are necessary in severe accident were assessed and will be further tested and/or analyzed. Such tests will consider severe accident conditions arised from hydrogen combustion.

Calculation of Evaporation From Fukushima Nuclear Power Plant Spent Fuel Pools

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Mirza M. Shah
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Spent Fuel ◽  
Decay Heat ◽  
Wide Range ◽  
Spent Fuel Pool ◽  
Safety Systems

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and postaccident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The postaccident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25% lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

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