Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents
One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600/sup 0/C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth.
- Research Organization:
- Oak Ridge National Lab., TN (USA)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 5798195
- Report Number(s):
- CONF-850570-6; ON: DE85012102
- Resource Relation:
- Conference: IAEA specialists' meeting on gas cooled reactors, Oak Ridge, TN, USA, 13 May 1985
- Country of Publication:
- United States
- Language:
- English
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21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
HTGR TYPE REACTORS
LOSS OF FLOW
AFTER-HEAT REMOVAL
HEATING
COMPUTER CODES
PRIMARY COOLANT CIRCUITS
SOILS
ACCIDENTS
COOLING SYSTEMS
ENERGY SYSTEMS
GAS COOLED REACTORS
GRAPHITE MODERATED REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
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220900* - Nuclear Reactor Technology- Reactor Safety
210300 - Power Reactors
Nonbreeding
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