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.
Skip Nav Destination
16th International Conference on Nuclear Engineering
May 11–15, 2008
Orlando, Florida, USA
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
0-7918-4817-5
PROCEEDINGS PAPER
Problems With Cooling of Overheated Fuel Channels in RBMK-Type Reactors
Eugenijus Uspuras,
Eugenijus Uspuras
Lithuanian Energy Institute, Kaunas, Lithuania
Search for other works by this author on:
Algirdas Kaliatka
Algirdas Kaliatka
Lithuanian Energy Institute, Kaunas, Lithuania
Search for other works by this author on:
Eugenijus Uspuras
Lithuanian Energy Institute, Kaunas, Lithuania
Algirdas Kaliatka
Lithuanian Energy Institute, Kaunas, Lithuania
Paper No:
ICONE16-48266, pp. 617-625; 9 pages
Published Online:
June 24, 2009
Citation
Uspuras, E, & Kaliatka, A. "Problems With Cooling of Overheated Fuel Channels in RBMK-Type Reactors." Proceedings of the 16th International Conference on Nuclear Engineering. Volume 4: Structural Integrity; Next Generation Systems; Safety and Security; Low Level Waste Management and Decommissioning; Near Term Deployment: Plant Designs, Licensing, Construction, Workforce and Public Acceptance. Orlando, Florida, USA. May 11–15, 2008. pp. 617-625. ASME. https://doi.org/10.1115/ICONE16-48266
Download citation file:
8
Views
Related Proceedings Papers
Related Articles
The Plant Feature and Performance of Double MS (Modular Simplified and Medium Small Reactor)
J. Eng. Gas Turbines Power (January,2010)
Quantitative and Qualitative Comparison of Light Water and Advanced Small Modular Reactors
ASME J of Nuclear Rad Sci (October,2015)
Related Chapters
PSA Level 2 — NPP Ringhals 2 (PSAM-0156)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
LOCA Frequencies Estimated from Operating Experience (PSAM-0282)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Modeling of SAMG Operator Actions in Level 2 PSA (PSAM-0164)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)