#Quan an vat aiya full#

Elevated-temperature testing is used to assess behavior under abnormal reactor situations. Results from elevated-temperature testing on samples from this element and a comparison with the EBR-II data are the subjects of this paper. A description of the experiment, as well as the FCCI results from as-irradiated samples, was reported by Tsai and Neimark. Element 181193 from the IFR-I experiment was destructively examined in Argonne National Laboratory`s alpha-gamma hot cell facility. The comparison would establish a link between the large EBR-II database of FCCI information to that of the prototypical power/temperature profile, which would be required for the licensing of the integral fast reactor (IFR). Analysis of various specimens from EBR-II experiments has shown that different combinations of linear power and cladding temperature significantly affect more » FCCI. One purpose of the experiment was to compare the fuel/cladding chemical interaction (FCCI) behavior between elements irradiated with a relatively flat power profile in the Experimental Breeder Reactor if (EBR-II) to full-length (91.44 cm of fuel) elements irradiated with a more prototypical power and temperature profile, as in the FFTF. The IFR-l experiment contained fuel elements with a U-1 wt% Pu-10 wt% Zr alloy in D9 cladding and was irradiated to 9.4 at.% in the Fast Flux Test Facility (FFTF).

#Quan an vat aiya full#

Hanford Engineering Development Laboratory (HEDL) will supply standard test article hardware for the 169 element assembly, will assemble the test article and irradiate the assembly to a goal exposure of 600 full power days (FPD). The completed fuel elements will be shipped to the Hanford Engineering Development Laboratory (HEDL) for wire wrapping and subassembly fabrication.

Metallic fuel slugs will be injection cast in three nominal compositions (U-10Zr, U-8Pu-10Zr and U-19Pu-10Zr) and fabricated more » into fuel elements at Argonne-West in Idaho. The test is part of the Integral Fast Reactor (IFR) fuels program conducted by Argonne National Laboratory in support of the Innovative Reactor Concepts Program sponsored by the US Department of Energy (DOE). The IFR-1 Test, designated as the AK-181 Test Assembly, will be the first irradiation test of wire-wrapped sodium-bonded metallic fuel elements (169) in the Fast Flux Test Facility (FFTF). The Test Design Description, Volume 1B, specifies the engineering requirements for the design and fabrication of the IFR-1 metallic fuel test. The elements conform to the requirements in MG-22, 'User`s Guide for the Irradiation of Experiments in the FTR.' « less This Test Design Description contains only the relevant fabrication data for the metal fuel elements built at Argonne. The metal fuel elements were fabricated at ANL-W and delivered to Westinghouse-Hanford for assembly into the test article. long depleted-uranium axial blanket slug (DU-10Zr) at each end. long, composed of three slugs and has one 6.5-in. The metal fuel elements contain U-10Zr cast slugs and have a fuel-smeared density of 75%. The elements conform to the requirements in MG-22, 'Users` Guide for the Irradiation of Experiments in the = F to 450 EFPD or until a cladding breach occurs. This Test Design Description contains relevant data on compositions, densities, dimensions and weights for the cast fuel slugs and completed fuel elements. the fuel elements were fabricated at ANL-W and delivered to Westinghouse-Hanford for wirewrapping and assembly into the test article. long depleted uranium axial blanket slug (DU-10Zr) was loaded at each end of the enriched zone. The enriched zone contains three slugs and is 36-in. The metal fuel elements have a fuel-smeared density of 75% and each contains five slugs. Three metal fuel alloys (U-10Zr, U-8Pu-10Zr) will be irradiated in D9 cladding tubes. One subassembly, containing 169 fuel elements, will be irradiated for 600 full power days to achieve 10 at.% burnup. The test is part of the Integral Fast Reactor (IFR) fuels program conducted by Argonne National Laboratory (ANL) in support of the Innovative Reactor Concepts Program sponsored by the US Department of Energy (DOE). The IFR-1 Test, designated as the AK-181 Test Assembly, will be the first irradiation test of wire wrapped, sodium-bonded metallic fuel elements in the Fast Flux Test Facility (FFTF).