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Title: FUEL CYCLE DEVELOPMENT PROGRAM. Quarterly Progress Report, January 1 to March 31, 1962

Abstract

The permanent shutdown of the Westinghouse Testing Reactor at the end of the quarter forced a revision in plans for programming the remainder of the irradiation of the two uninstrumented capsules whose testing was in progress. The minimum estimated burnup at this point, based on hot cell data obtained from the pilot capsule, was 13,100 MW-d/ton U. It was decided to continue the testing of only one capsule in another reactor until the original goal of 20,000 MW-d/ton U is reached. The irradiation of the second capsule is to be terminated so that it can serve as a control. Fabrication was initiated on enriched UO/sub 2/ pellets for incorporation in full scale fuel rods to be irradiated in the Vallecitos Boiling Water Reactor. A wet nitrogen pyrohydrolysis step in conjunction with oxidationreduction cycling is being used to attain a satisfactory density exceeding 95% of theoretical at 1150 deg C. Apparatus and procedures being used for measurement of thermal conductivity and thermal expansion of sintered and cast uranium carbide are described. The coefficient of linear thermal expansion for a single specimen of 4.37 wt.% carbon sintered uranium carbide was determined to be 11.8 x 10/sup -6/ mm/mm- deg C, while thatmore » for a cast uranium carbide specimen containing 4.73 wt.% carbon was 10.6 x 10/sup -6/ mm/mm- deg C over the range from room temperature to l000 deg C. Hardness data for 4.4 and 4.8 plus or minus 0.1 wt.% carbon sintered and cast uranium carbide and for 5.2 wt.% carbon cast material are also presented. Irradiation test specimens previously prepared were given a pre-irradiation examination and were encapsulated. Three capsules, two using a sodium bond between fuel and container and one using an interference fit between fuel and container, were loaded in the MTR and appear to be operating under the required conditions. (auth)« less

Publication Date:
Research Org.:
United Nuclear Corp. Development Div., White Plains, N.Y.
Sponsoring Org.:
USDOE
OSTI Identifier:
4772538
Report Number(s):
UNC-5015
NSA Number:
NSA-16-030026
DOE Contract Number:
AT(30-1)-2374
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-62
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; BONDING; BREEDING; BURNUP; CAPSULES; CARBON; CASTING; DENSITY; EXPANSION; FABRICATION; FUEL CANS; FUEL CYCLE; FUELS; HARDNESS; HIGH TEMPERATURE; INSTRUMENTS; IRRADIATION; LABORATORY EQUIPMENT; MATERIALS TESTING; MTR; NITROGEN; PELLETS; PYROLYSIS; QUANTITY RATIO; REACTORS; REPROCESSING; SINTERING; SODIUM; TESTING; THERMAL CONDUCTIVITY; URANIUM CARBIDES; URANIUM DIOXIDE; VBWR; WTR

Citation Formats

None. FUEL CYCLE DEVELOPMENT PROGRAM. Quarterly Progress Report, January 1 to March 31, 1962. United States: N. p., 1962. Web. doi:10.2172/4772538.
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None. FUEL CYCLE DEVELOPMENT PROGRAM. Quarterly Progress Report, January 1 to March 31, 1962. United States. doi:10.2172/4772538.
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None. Fri . "FUEL CYCLE DEVELOPMENT PROGRAM. Quarterly Progress Report, January 1 to March 31, 1962". United States. doi:10.2172/4772538. https://www.osti.gov/servlets/purl/4772538.
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@article{osti_4772538,
title = {FUEL CYCLE DEVELOPMENT PROGRAM. Quarterly Progress Report, January 1 to March 31, 1962},
author = {None},
abstractNote = {The permanent shutdown of the Westinghouse Testing Reactor at the end of the quarter forced a revision in plans for programming the remainder of the irradiation of the two uninstrumented capsules whose testing was in progress. The minimum estimated burnup at this point, based on hot cell data obtained from the pilot capsule, was 13,100 MW-d/ton U. It was decided to continue the testing of only one capsule in another reactor until the original goal of 20,000 MW-d/ton U is reached. The irradiation of the second capsule is to be terminated so that it can serve as a control. Fabrication was initiated on enriched UO/sub 2/ pellets for incorporation in full scale fuel rods to be irradiated in the Vallecitos Boiling Water Reactor. A wet nitrogen pyrohydrolysis step in conjunction with oxidationreduction cycling is being used to attain a satisfactory density exceeding 95% of theoretical at 1150 deg C. Apparatus and procedures being used for measurement of thermal conductivity and thermal expansion of sintered and cast uranium carbide are described. The coefficient of linear thermal expansion for a single specimen of 4.37 wt.% carbon sintered uranium carbide was determined to be 11.8 x 10/sup -6/ mm/mm- deg C, while that for a cast uranium carbide specimen containing 4.73 wt.% carbon was 10.6 x 10/sup -6/ mm/mm- deg C over the range from room temperature to l000 deg C. Hardness data for 4.4 and 4.8 plus or minus 0.1 wt.% carbon sintered and cast uranium carbide and for 5.2 wt.% carbon cast material are also presented. Irradiation test specimens previously prepared were given a pre-irradiation examination and were encapsulated. Three capsules, two using a sodium bond between fuel and container and one using an interference fit between fuel and container, were loaded in the MTR and appear to be operating under the required conditions. (auth)},
doi = {10.2172/4772538},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Jun 01 00:00:00 EDT 1962},
month = {Fri Jun 01 00:00:00 EDT 1962}
}
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Technical Report:
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DOI:  10.2172/4772538
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  • Activities in an integrated investigation in the VBWR and other facilities to improve the technological limits of boiling reactors are reported. In work on advanced fuel power tests, reduction of VBWR stability test data is almost complete. Information is also included on the irradiation status of fuel assemblies and operation of defected fuel tests. Results of UO/sub 2/ conductivity tests in the center melting calibration assembly are reported. In research on heat transfer and fluid dynamics, a seven-rod test section was fabricated for investigation of differences between single- and multi-rod burnout heat flux. Measurements of fuel isotopic composition indicate thatmore » the capture of neutrons by U/sup 238/ to form Pu/sup 239/ may have been underestimated. (J.R.D.)« less

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  • The laboratory scale development phase of a process for the fabrication of UO/sub 2/ pellets based on low temperature sintering in nitrogen was completed. Work was initiated on the preparation and encapsulation of enriched UO/sub 2/ pellets for irradiation testing. Further investigation of the effect of fluoride additions on the sintering behavior of UO/sub 2/ con firmed that oxidation-reduction cycling restored the oxide sinterability lost as a result of fluoridation. A series of sintering experiments were performed with the objective of producing high density UO/sub 2/ pellets in excess of 98 percent theoretical density. Modification of the oxidation-reduction activation treatmentmore » could produce such densities. It also appeared that sintering temperatures as low as 100 deg C can be used to prepare pellets of the design density of 95 percent theoretical density. The propane reaction was further studied to develop techniques for preparing uranium carbide powder to a given total carbon specification. A comparison was made between the use of unreacted UO/sub 2/-graphite and reacted UO/sub 2/-graphite charges to the arc skull furnace. The reacted charge gave the best results. The determination of the physical properties of uranium carbide as a function of composition and fabricational variables was initiated. The thermal expansion of cast 4.85 wt% C uranium carbide was determined at 12.55 x 10 /sup -6/ in./in./deg C. (auth)« less

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