HTGR Knowledge Base

Conference Article: ORNL analyses of AVR performance and safety

Cleveland, J.C. (Oak Ridge National Lab., TN (USA))

Abstract

Because of the high interest in modular High Temperature Reactor performance and safety, a cooperative project has been established involving the Oak Ridge National Laboratory (ORNL), Arbeitsgemeinschaft Versuchs Reaktor GmbH (AVR), and Kernforschungsanlage Juelich GmbH in reactor physics, performance and safety. This project has been established within the frame of the HTR Implementing Agreement between the U.S. Department of Energy and the Ministry of Research and Technology of the Federal Republic of Germany. The effort focuses on examination of AVR neutronic and thermal hydraulic behavior with objectives of further insights into pebble bed reactor behavior. An independent analysis by ORNL of the thermal response of the AVR to a hypothetical depressurized core heatup accident has confirmed analyses performed by AVR and KFA staff in consideration of potential testing of the AVR under depressurized heatup conditions. ORNL results show that the cooler regions of the core and reflector provides an ''early'' heat sink limiting the maximum fuel temperature to below 1350 deg. C. The fuel remains above normal temperature for four to five days and during this period, the maximum fuel temperature is not determined by the rate of heat dissipation through the outer vessel. The effect on inner vessel temperature of uncertainties in the thermal properties of the first biological shield are also investigated. ORNL is also analyzing the response of the AVR to tests involving coolant flow reductions without control rod motion. Results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. The results reported here illustrate the importance of safety features, such as the ability to remove afterheat through the vessel wall, the ability to limit temperature changes due to the large core heat capacity, and the close coupling of core power generation and heat removal capability

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key words: after-heat removal; avr reactor; design basis accidents; loss of flow; reactivity coefficients; temperature distribution; very high temperature; accidents; enriched uranium reactors; gas cooled reactors; graphite moderated reactors; helium cooled reactors; homogeneous reactors; htgr type reactors; pebble bed reactors; power reactors; reactor accidents; reactors; solid homogeneous reactors; thermal reactors; thorium reactors
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:73-84