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Title: NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT

Abstract

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)

Authors:
; ;
Publication Date:
Research Org.:
Battelle Memorial Inst., Columbus, OH (United States)
Sponsoring Org.:
USDOE
OSTI Identifier:
4170636
Report Number(s):
BMI-1314
NSA Number:
NSA-14-008280
DOE Contract Number:  
W-7405-ENG-92
Resource Type:
Technical Report
Resource Relation:
Other Information: Decl. Dec. 3, 1959. Orig. Receipt Date: 31-DEC-60
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; BRR; CYLINDERS; DISTRIBUTION; FUEL CANS; FUEL ELEMENTS; GAS COOLANT; GAS FLOW; IN PILE LOOPS; MEASURED VALUES; NEUTRON FLUX; POWER; POWER PLANTS; QUANTITY RATIO; STAINLESS STEELS; TEMPERATURE; URANIUM OXIDES

Citation Formats

Anno, J. N., Fairand, B. P., and Chastain, Jr., J. W. NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT. United States: N. p., 1959. Web. doi:10.2172/4170636.
Anno, J. N., Fairand, B. P., & Chastain, Jr., J. W. NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT. United States. https://doi.org/10.2172/4170636
Anno, J. N., Fairand, B. P., and Chastain, Jr., J. W. 1959. "NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT". United States. https://doi.org/10.2172/4170636. https://www.osti.gov/servlets/purl/4170636.
@article{osti_4170636,
title = {NEUTRON-FLUX MEASUREMENTS IN A CONCENTRIC-CYLINDER FUEL ELEMENT},
author = {Anno, J. N. and Fairand, B. P. and Chastain, Jr., J. W.},
abstractNote = {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)},
doi = {10.2172/4170636},
url = {https://www.osti.gov/biblio/4170636}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Jan 29 00:00:00 EST 1959},
month = {Thu Jan 29 00:00:00 EST 1959}
}