Atr Fuel Materials Development Irradiation Results. Part Ii

Abstract

This report w88 prepared 88 an account of Government sponsored work. Neither the United States, nor the Commission. nor any person acting on behalf of the Commisslon:A. Makes any warranty o r representation, expressed or implied, with respect to the accuracy. cornpleteneea, or usefulness of the Information contained in this report. or that the use of any information, apparatus. method. or process dlsclosed in this report may not infringe privately owned rights: or B. Assumes any liabilitles with respect to the use of. or for damages resulting from the use of any information. apparatue. method, or process dlscloeed In this report.As used in the above. "person acting on behalf of the Commisaion" includes any employee or contractor of the Commission, o r employee of such contractor, to the extent that such employee or contractor of the Commisslon, or employee of such contractor prepares, disseminates, or provides B C C~B B to, any information pursuant to his employment or contract with the Commission. o r hia employment with such contractor. Post-irradiation anneal experiments on plates containing the nominal ATR fuel and poison loadings showedthe threshold temperature for breakaway swelling t o be 750°F for the U308 fueled plates and 800°F for UAl3 plates which had a fission density of 15 x 1020 f/cc. The breakaway swelling temperature for a Nb-coated U02-Zr02 plate with the same U-235 and B-10 content was 500°F with a fission density of 11 x 1020. PH I 111 PSiii S U M M A R YThe work described in this report is a continuation of an irradiation program conducted upon fuel materials of interest to high-performance test reactors such a s the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR). A series of proof-tests conducted previously showed that blisters formed on sample fuel plates at fission densities above about 8 x 1020 fissions per cubic centimeter (f/cc) while operating at temperatures which were not known with satisfactory precision. A lcop facility (known a s PAED-G-12 ETR Water Loop) was de signed and constructed in which more carefully controlled irradiation experiments were executed. The operating conditions that could be achieved corresponded to the locations in the reactors which were operating at the maximum nominal power density (ie, not including off -normal contributions known a s hot spot-hot channel factors).The fuel materials tested, the maximum nominal cladding temperature and surface heat flux, and the average fission densities obtained are as shown in the following tabulation: t o 17All the fuel samples were clad with 6061-0 aluminum.No evidence of blister formation o r other deleterious effects attributable to fuel material stability was observed during the irradiations except that one of the U 0 2 fueled plates became severely warped; the cause for this behavior is believed to be the poor distribution of fuel within this particular plate.Measurements of the cladding temperature of instrumentedplate s during the course of the irradiations showed that the corrosion of th...

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

“…This work focuses on the Mark VII fuel design. Information related to legacy fuel forms is considered only as needed to supplement the Mark VII fuel literature [13][14][15]. An ongoing effort is under way to convert the ATR core to low-enriched uranium (LEU) [16,17].…”

“…A later study published in 1958 reported that the stable uranium aluminides include UAl 2 , UAl 3 , and UAl 4 [20]. The discrepancy may be related to crystalline defects in the UAl 4 lattice, which displace U atoms and produce a U:Al stoichiometry closer to 1:5 [14]. Uranium aluminides are prepared by vacuum induction melting, arc melting, casting, or reaction of uranium hydride with Al [11,18].…”

Irradiation Testing of Fuel for the Mark I Core of the Argonne Advanced Research Reactor.

“…In view of these new parameters in fuel element operation, it was believed that an irradiation test program was worthwhile. A similar program directed toward the needs of the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) had just been completed [2], and space was available in the irradiation facility used for this test. The ATR and HFIR fuel elements use aluminum clad dispersion-type fuel elements.…”