A modular gas-cooled reactor design with a thermal output of 600MWt and a core exit temperature of 950° C has been designed by the Korea Atomic Energy Research Institute based on the GT-MHR reactor concept which adopts a prismatic core. A sensitivity study on the transient plant behavior during a postulated depressurized LOFC accident concurrent with the failure of the RCCS was performed. In the transient analysis, the GAMMA+ code which can handle multi-dimensional, multi-component problem was used. The RCCS is a passive system which is very reliable and supplies a significant heat removal mechanism during abnormal conditions in a GCR. To investigate the safety characteristics of a GCR under the one of the worst accidental scenarios, a simultaneous failure of the RCCS with a depressurized LOFC was assumed. The thermal behavior of the reactor system was analyzed in various conditions. It is found that the maximum temperature of the reactor fuel compact could exceed 1600° C at about 50 hours at the condition of a depressurized LOFC with a failure of the RCCS. A problem with the structural integrity of the reactor pressure vessel could also be a critical factor. The insulation of a reactor cavity wall serves as a dominant obstacle against a heat transfer from the reactor vessel to the surrounding ground when the RCCS fails to operate. Without insulation material on the reactor cavity wall, the gradients of the increasing rate of the maximum temperature diminish and the peak values decrease. The maximum temperatures of the fuel compact and the reactor vessel are less sensitive to the concrete and surrounding soil properties, those are the thermal conductivity and volumetric heat capacity, when the insulation material is used. The uncertainties in the properties of the concrete and the surrounding soil become significant without an insulation material in the cavity. To improve the safety of a modular GCR, more effective and feasible heat removal mechanism need to be devised based on the comprehensions on the heat transfer characteristics.
Skip Nav Destination
Fourth International Topical Meeting on High Temperature Reactor Technology
September 28–October 1, 2008
Washington, DC, USA
Conference Sponsors:
- ASME
ISBN:
978-0-7918-4855-5
PROCEEDINGS PAPER
Sensitivity Study on Depressurized LOFC Accidents With a Failure of the RCCS in a Modular Gas-Cooled Reactor
Sang Jun Ha,
Sang Jun Ha
Korea Electric Power Research Institute, Daejeon, South Korea
Search for other works by this author on:
Seyun Kim,
Seyun Kim
Korea Electric Power Research Institute, Daejeon, South Korea
Search for other works by this author on:
Nam-Il Tak,
Nam-Il Tak
Korea Atomic Energy Research Institute, Daejeon, South Korea
Search for other works by this author on:
Hong Sik Lim
Hong Sik Lim
Korea Atomic Energy Research Institute, Daejeon, South Korea
Search for other works by this author on:
Sang Jun Ha
Korea Electric Power Research Institute, Daejeon, South Korea
Seyun Kim
Korea Electric Power Research Institute, Daejeon, South Korea
Nam-Il Tak
Korea Atomic Energy Research Institute, Daejeon, South Korea
Hong Sik Lim
Korea Atomic Energy Research Institute, Daejeon, South Korea
Paper No:
HTR2008-58307, pp. 233-240; 8 pages
Published Online:
July 1, 2009
Citation
Ha, SJ, Kim, S, Tak, N, & Lim, HS. "Sensitivity Study on Depressurized LOFC Accidents With a Failure of the RCCS in a Modular Gas-Cooled Reactor." Proceedings of the Fourth International Topical Meeting on High Temperature Reactor Technology. Fourth International Topical Meeting on High Temperature Reactor Technology, Volume 2. Washington, DC, USA. September 28–October 1, 2008. pp. 233-240. ASME. https://doi.org/10.1115/HTR2008-58307
Download citation file:
6
Views
0
Citations
Related Proceedings Papers
Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs
ICONE20-POWER2012
Related Articles
The Fabulous Nuclear Odyssey of Belgium
J. Pressure Vessel Technol (June,2009)
Computational Fluid Dynamic Analysis of the ESFR Reactor Pit Cooling System in Case of Sodium Leakage
ASME J of Nuclear Rad Sci (April,2022)
COBRA-TF Simulation of DNB Response During Reactivity-Initiated Accidents Using the NSRR Pulse Irradiation Experiments
ASME J of Nuclear Rad Sci (July,2016)
Related Chapters
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
PSA Level 2 — NPP Ringhals 2 (PSAM-0156)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
Source Term Assessments in PSA Level 2 for the Outage Period (PSAM-0168)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)