ICONE11-36108 SEVERE ACCIDENT MANAGEMENT MEASURES IN VVER-91/99

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.

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Study on Severe Accident Management Measures for the Piping System in Nuclear Power Plant using Vibration Control Techniques

The Proceedings of the Dynamics & Design Conference ◽

10.1299/jsmedmc.2017.203 ◽

2017 ◽

Vol 2017 (0) ◽

pp. 203

Author(s):

Osamu FURUYA ◽

Sunggook CHO ◽

Hiroshi KURABAYASHI

Keyword(s):

Power Plant ◽

Nuclear Power Plant ◽

Vibration Control ◽

Nuclear Power ◽

Severe Accident ◽

Piping System ◽

Control Techniques ◽

Management Measures ◽

Accident Management

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

Volume 2: Plant Systems, Structures, and Components; Safety and Security; Next Generation Systems; Heat Exchangers and Cooling Systems ◽

10.1115/icone20-power2012-54568 ◽

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.

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Severe accident management measures for a generic German PWR. Part II: Small-break loss-of-coolant accident

Annals of Nuclear Energy ◽

10.1016/j.anucene.2018.08.017 ◽

2018 ◽

Vol 122 ◽

pp. 280-296 ◽

Cited By ~ 3

Author(s):

M. Jobst ◽

P. Wilhelm ◽

Y. Kozmenkov ◽

S. Kliem

Keyword(s):

Severe Accident ◽

Loss Of Coolant Accident ◽

Management Measures ◽

Accident Management ◽

Loss Of Coolant

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

Science and Technology of Nuclear Installations ◽

10.1155/2013/598975 ◽

2013 ◽

Vol 2013 ◽

pp. 1-11 ◽

Cited By ~ 2

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.

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