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Title:
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Afterheat removal from a helium reactor under accident conditions: CFD calculations for the code-to-code benchmark analyses on the thermal behaviour for the gas turbine modular helium reactor
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Author(s):
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Published by:
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Publication date:
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ECN
NUCLEAIR
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1998
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ECN report number:
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Document type:
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ECN-RX--97-066
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Article (scientific)
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Number of pages:
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11
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Published in: To be published in the proceedings of the NEA workshop on 'High temperature engineering research facilities and experiments' (12 (), , , Vol., p.-.
Abstract:
The International Atomic Energy Agency (IAEA) Co-ordinated ResearchProgramme (CRP) on 'Heat Transport and Afterheat Removal for Gas Cooled
Reactors under Accident Conditions' has organised benchmark analyses to
support verification and validation of analytical tools used by the
participants to predict the thermal behaviour of advanced gas cooled reactors
during accidents. One of thew benchmark analyses concerns the code-to-code
analysis of the Gas Turbine Modular Helium Reactor (GT-MHR) plutonium burner
accidents. The GT-MHR is a passive safe, helium cooled, graphite moderated,
advanced reactor system with a thermal power of 600 MW that is based on
existing technology. The GT-MHR can also be fuelled with plutonium. If the
main helium cooling and the auxiliary shut-down cooling systems fail or
become unavailable, the core afterheat is removed by radiation and convection
inside the reactor vessel and the reactor cavity to the Reactor Cavity
Cooling System (RCCS). The objective of the RCCS is to serve as an ultimate
heat sink, ensuring the thermal integrity of the core, vessel and critical
equipment within the reactor cavity for the entire spectrum of postulated
accident sequences. This paper describes the heat transport inside the
reactor core to the RCCS. For this purpose, the heat transfer mechanisms as
well as the flow patterns inside the core, the reactor pressure vessel, and
the cavity have been calculated by the Computational Fluid Dynamics (CFD)
code CFX-F3D). The behaviour of the RCCS itself is not described. One
calculation considers the full power operation, while two calculations
consider Loss Of Forced Convection (LOFC) accidents, one at pressurised
conditions and the other depressurised conditions. The heat transfer from the
reactor vessel to the environment under normal operation conditions is 2.64
MW. The highest temperature in the core is 1222K, and the average core
temperature is 1075K. The highest reactor vessel temperature is 679K. The
highest temperatures are reached during a depres- surised LOFC accident the
highest core temperature is 1644K, and the highest vessel wall temperature is
736K. The highest vessel wall temperatures occur near the mid plane of the
core for the depressurised LOFC, and in the upper part of the reactor vessel
for the pressurised LOFC. The heat loss of the reactor vessel during
accidents is about 2.2 MW. A comparison between results of the calculations
performed by other benchmark participants cannot be made yet, because these
calculations are still in progress. 7 refs.
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