Transient void fraction measurement in simulating subchannels of PWR fuel assemblies

Journal Article · Sat Dec 31 04:00:00 EST 1994 · Transactions of the American Nuclear Society

OSTI ID:89329

Hori, Keiichi; Akiyama, Yoshiei; Miyazaki, Keiji; Kurosu, Tatsuo; Sugiyama, Shigekazu

The void fraction in a pressurized water reactor (PWR) core is negligible under normal operation conditions but becomes significant in high-powered fuel assemblies under anticipated transients or accident conditions. The void generation in the fuel assemblies provides redistribution of the coolant flow throughout the core and the void reactivity feed-back mechanism reduces the fission power. Therefore, void behavior in the fuel assemblies is one of the most important factors from the viewpoint of reactor safety even in a PWR nuclear power plant. The void fraction measurement tests for PWR fuel assemblies have been performed since 1987 under the sponsorship of the Ministry of International Trade and Industries as one of Japan`s national projects. One of the major objectives is to assess the void prediction method. In these tests, the single-channel transient test was performed in advance of the rod bundle test in 1992, and the obtained transient void data have been processed and used to assess the void prediction method.

• OSTI ID: 89329
• Report Number(s): CONF-941102--
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 71; ISSN 0003-018X; ISSN TANSAO
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
DENSITOMETERS
FUEL ASSEMBLIES
HYDRAULICS
MEASURING METHODS
PWR TYPE REACTORS
TWO-PHASE FLOW
VOID FRACTION