Prediction of heat transfer for a supercritical water test with a four pin fuel bundle
- RWE Power AG, Essen (Germany)
- Forschungszentrum Karlsruhe GmbH, Institute for Nuclear and Energy Technologies, PO Box 3640, D-76021 Karlsruhe (Germany)
- University of Stuttgart (Germany)
As a next step to validate prediction methods for core design of a Supercritical Water Cooled Reactor, a small, electrically heated fuel bundle with 4 pins is planned to be tested. This paper summarizes first heat transfer predictions for such a test, which were performed based on supercritical and subcritical sub-channel analyses. For heat transfer under supercritical pressure conditions, the sub-channel code STAFAS has been applied, which had been tested successfully already for a supercritical water reactor design. Design studies with different assembly box sizes at a given pin diameter and pitch have been performed to optimize the coolant temperature distribution. With a fuel pin outer diameter of 10 mm and a pitch to diameter ratio of 1.15, an optimum inner width of the assembly box was determined to be 24 mm. Coolant and cladding surface temperatures to be expected at subcritical pressure conditions have been predicted with the sub-channel code MATRA. As, different from typical PWR or BWR conditions, a dryout has been foreseen for the tests, this code had to be extended to include suitable dryout criteria as well as post dryout heat transfer correlations at higher enthalpies and pressures. Different from PWR or BWR design, the cladding surface temperature of fuel pins in supercritical water reactors can vary significantly around the circumference of each pin, causing bending towards its hotter side which, in turn, can cause additional sub-channel heat-up and thus additional thermal bending of the pin. To avoid a thermal instability by this effect, a sensitivity study with respect to thermal bending of fuel pins has been performed, which determines the minimum number of grid spacers needed for this test. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 21021031
- Resource Relation:
- Conference: 2006 International congress on advances in nuclear power plants - ICAPP'06, Reno - Nevada (United States), 4-8 Jun 2006; Other Information: Country of input: France; 20 refs; Related Information: In: Proceedings of the 2006 international congress on advances in nuclear power plants - ICAPP'06, 2734 pages.
- Country of Publication:
- United States
- Language:
- English
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