Proof test on thermal-hydraulic design reliability of a BWR fuel assembly

Conference · Thu Jan 01 04:00:00 EST 1987 · Trans. Am. Nucl. Soc.; (United States)

OSTI ID:6810358

Akiyama, M; Inoue, A; Ohishi, M; Aoki, T; Kagawa, T; Yoshimura, K

A series of critical power tests for boiling water reactor (BWR) fuel assemblies was carried out using a test facility that has a very different system from that of General Electric (GE). It was reported in 1977 that although the FRIGG data were systematically approx. 5% higher than those from ATLAS, there was no evidence of any test facility data bias when the combined measurement errors were considered. The purpose of this test is to investigate qualitatively the difference between the data obtained from the different facility and the value calculated by the GEXL correlation under various conditions. The test facility consists mainly of two loops - a recirculation loop and a cooling loop. The recirculation loop consists of circulating pumps, a preheater, and a test section. The test facility can simulate steady and transient thermal-hydraulic conditions at a BWR core by computer control. The test assemblies, which have an 8 x 8 full-scale fuel configuration, are heated by direct current to simulate nuclear heating profiles. The obtained critical power data are in good agreement with the GEXL correlation predicted value (ECPR = 1.001, standard deviation = 3.4%) despite testing in a dissimilar loop, differently designed test assemblies compared to GE ATLAS, and the test condition of a wide range of parameters.

• Research Organization: Univ. of Tokyo (Japan)
• OSTI ID: 6810358
• Report Number(s): CONF-8711195-
• Conference Information: Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 55
• Country of Publication: United States
• Language: English

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BWR TYPE REACTORS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLUID MECHANICS
FUEL ASSEMBLIES
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REACTOR COOLING SYSTEMS
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