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Conference Article: Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heat up accidents

Harrington, R.M.; Ball, S.J. (Oak Ridge National Lab., TN (USA))

Abstract

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 at Oak Ridge National Laboratory under the auspices of the U.S. Nuclear Regulatory Commission's HTGR Research Program, to test this important 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. The code is capable of predicting the peak fuel, reflector, and reactor vessel temperatures reached following permanent loss of forced primary coolant circulation accidents with or without concomitant loss of steam generator cooling water flow or accidental depressurization of the helium primary coolant. 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 deg. 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. This model was used 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. The simulation 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. Soil properties at some sites may not be suitable.

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key words: after-heat removal; computer calculations; depressurization; heat flux; htgr type reactors; loss of flow; primary coolant circuits; steam generators; temperature distribution; thermal conduction; very high temperature; accidents; boilers; cooling systems; energy transfer; gas cooled reactors; graphite moderated reactors; heat transfer; reactor accidents; reactor components; reactor cooling systems; reactors; vapor generators
Reference:
Specialists' meeting on safety and accident analysis for gas-cooled reactors. Oak Ridge, TN (USA). 13-15 May 1985
International Atomic Energy Agency, Vienna (Austria)
IAEA-TECDOC--358, pp:273-282