Effect of rod bow to partial closure on critical heat flux in PWR fuel assembly

Journal Article · Fri Jul 01 04:00:00 EDT 1983 · Am. Soc. Mech. Eng., (Pap.); (United States)

OSTI ID:6987308

Macbuff, R B; Fighetti, C F

The effects of partial closure due to bowed rods on critical heat flux (CHF) in a pressurized water reactor rod bundle were evaluated by conducting tests in an electrically heated test section. The test section consisted of a 5x5 square rod array with 24 heated rods of 9.14 mm (0.360 in) diameter, each with a heated length of 3.66 m (12 ft) with Exxon Nuclear Company spacer grids on a 0.521 m (20.5 in) pitch. The central rod was a 12.2 mm (0.480 in) diameter unheated guide tube. The gap between two centrally located high powered rods was reduced 71% from nominal. The bow was approximately mid-span between two grids in the region in which CHF was observed in unbowed test sections. The results of these tests indicate a reduction in CHF of approximately 5% at 2400 psia and no reduction at 2000 psia when test repeatability is taken into consideration. The reduction in CHF at 2400 psia is substantially smaller than that inferred by linear interpolation of previously reported test results at 50 and 100% reduction of spacing. The local nature of CHF reduction due to rod bow was confirmed.

• Research Organization: Exxon Nuclear Co., Inc Richland, WA
• OSTI ID: 6987308
• Journal Information: Am. Soc. Mech. Eng., (Pap.); (United States), Journal Name: Am. Soc. Mech. Eng., (Pap.); (United States) Vol. 83-HT-105; ISSN ASMSA
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200* -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
CRITICAL HEAT FLUX
DATA
DEFORMATION
EXPERIMENTAL DATA
FUEL ELEMENTS
FUEL RODS
HEAT FLUX
INFORMATION
NUMERICAL DATA
PERFORMANCE TESTING
PRESSURE DEPENDENCE
PWR TYPE REACTORS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
SAFETY
TESTING
WATER COOLED REACTORS
WATER MODERATED REACTORS