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

Creation Information

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

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Description

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)

Physical Description

Pages: 22

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NTIS

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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

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

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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 April 20, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.