NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT

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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 ... continued below

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Pages: 22

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Anno, J.N.; Fairand, B.P. & Chastain, J.W. Jr. January 29, 1959.

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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)

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Pages: 22

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  • Other Information: Decl. Dec. 3, 1959. Orig. Receipt Date: 31-DEC-60

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  • Report No.: BMI-1314
  • Grant Number: W-7405-ENG-92
  • Office of Scientific & Technical Information Report Number: 4170636
  • Archival Resource Key: ark:/67531/metadc870540

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Office of Scientific & Technical Information Technical Reports

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  • January 29, 1959

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  • Sept. 16, 2016, 12:32 a.m.

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  • Nov. 30, 2016, 12:47 p.m.

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Anno, J.N.; Fairand, B.P. & Chastain, J.W. Jr. NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT, report, January 29, 1959; Columbus, Ohio. (digital.library.unt.edu/ark:/67531/metadc870540/: accessed November 14, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.