NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT
Neutron-flux measurements in a concentric-cylinder fuel element were made in a gas-cooled in-pile loop operated adjacent to the core of the BRR. The fuel element comprised four concentric fuel cylinders. Each fuel annulus (outside diameters- 1.248 1.018, 0.810 and 0.590 in.) consisted of a 0.031-in.- thick core of UO/sub 2/ dispersed in type 347 stainless steel and clad on each side with 0.007 in. of typee 318 stainless steel. The element was 24 in. long and the total uranium-235 content was approximately 192 g. Radial, vertical, and peripheral flux distributions were studied. The vertical flux profile was cosine- shaped with a peak-to-average ratio of 1.26. The peripheral variation around the loop wall could also be fitted to a cosine curve (with a peak-to-average ratio of 1.10). The average radial flux depression from the outer fuel cylinder to the center of the element was a factor of 2.14. Power generation in the element calculated from flux measurements agreed to within 10% with the power generated by measuring gas now rate and temperarure rise across the fuel element. The ratio of peak-to-average power density was found to be 1.75. (auth)
- Research Organization:
- Battelle Memorial Inst., Columbus, Ohio
- Sponsoring Organization:
- USDOE
- DOE Contract Number:
- W-7405-ENG-92
- NSA Number:
- NSA-14-008280
- OSTI ID:
- 4170636
- Report Number(s):
- BMI-1314
- Country of Publication:
- United States
- Language:
- English
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