NEUTRON-FLUX MEASUREMENTS IN A FLAT PLATE FUEL ELEMENT
Abstract
The equipment and experiments performed to measure the thermal-neutron- flux distribution in a fuel assembly of an experimental loop mock-up of a gas:cooled reactor at the Battelle Research Reactor (BRR) are described. The loop was located adjacent to the core of the BRR and contained one fuel assermbly composed of seven flat fuel plates each containing approximately 29.5 g of U/sup 235/. The plates consisted of a core 0.050 in. thick of UO/sub 2/ dispersed in Type 347 stainless steel and clad on each side with 0.005 in. of Type 347 stainless steel. The measurements showed that with the present design of the loop system an average thermal-neutron flux of 4.09 x 1O/sup 12/ neutrons/(cm/sup 2/)(sec) or a power generation of 45 kw in the assembly can be conveniently obtained. The ratio of the peak thermal-neutron flux to average thermal flux in the entire element was found to be 1.87. At any horizontal cross section, thermal-flux depression from the edge of the element to the center of less than a factor of two was observed for the final loopcore arrangement. (auth)
- Authors:
- Publication Date:
- Research Org.:
- Battelle Memorial Inst., Columbus, Ohio
- Sponsoring Org.:
- USDOE
- OSTI Identifier:
- 4170970
- Report Number(s):
- BMI-1231
- NSA Number:
- NSA-14-008239
- 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; COOLANT LOOPS; FUEL ELEMENTS; GAS COOLANT; MEASURED VALUES; MOCKUP; NEUTRON FLUX; PLATES; REACTORS; STAINLESS STEELS; THERMAL NEUTRONS; URANIUM 235; URANIUM OXIDES
Citation Formats
Morgan, W.R., Anno, J.N. Jr., and Chastain, J.W. Jr. NEUTRON-FLUX MEASUREMENTS IN A FLAT PLATE FUEL ELEMENT. United States: N. p., 1959.
Web. doi:10.2172/4170970.
Morgan, W.R., Anno, J.N. Jr., & Chastain, J.W. Jr. NEUTRON-FLUX MEASUREMENTS IN A FLAT PLATE FUEL ELEMENT. United States. doi:10.2172/4170970.
Morgan, W.R., Anno, J.N. Jr., and Chastain, J.W. Jr. Thu .
"NEUTRON-FLUX MEASUREMENTS IN A FLAT PLATE FUEL ELEMENT". United States.
doi:10.2172/4170970. https://www.osti.gov/servlets/purl/4170970.
@article{osti_4170970,
title = {NEUTRON-FLUX MEASUREMENTS IN A FLAT PLATE FUEL ELEMENT},
author = {Morgan, W.R. and Anno, J.N. Jr. and Chastain, J.W. Jr.},
abstractNote = {The equipment and experiments performed to measure the thermal-neutron- flux distribution in a fuel assembly of an experimental loop mock-up of a gas:cooled reactor at the Battelle Research Reactor (BRR) are described. The loop was located adjacent to the core of the BRR and contained one fuel assermbly composed of seven flat fuel plates each containing approximately 29.5 g of U/sup 235/. The plates consisted of a core 0.050 in. thick of UO/sub 2/ dispersed in Type 347 stainless steel and clad on each side with 0.005 in. of Type 347 stainless steel. The measurements showed that with the present design of the loop system an average thermal-neutron flux of 4.09 x 1O/sup 12/ neutrons/(cm/sup 2/)(sec) or a power generation of 45 kw in the assembly can be conveniently obtained. The ratio of the peak thermal-neutron flux to average thermal flux in the entire element was found to be 1.87. At any horizontal cross section, thermal-flux depression from the edge of the element to the center of less than a factor of two was observed for the final loopcore arrangement. (auth)},
doi = {10.2172/4170970},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Thu Dec 03 00:00:00 EST 1959},
month = {Thu Dec 03 00:00:00 EST 1959}
}
-
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 withmore » -
SOME MEASUREMENTS OF THE NEUTRON FLUX IN THE SPENT FUEL ELEMENT IRRADIATION POND AT HARWELL
Indium foil activation showed the neutron fluxes in the assembly to be very small and to be the result of the ( gamma ,n) reaction on deuterium by 2.5- Mev gammas emitted by La/sup 140/. Correlation of neutron and gamma field showed that the strongest fields obtainable will not induce significant activity in irradiated samples. (auth)