Fuels for research and test reactors, status review: July 1982

Abstract

U0 2 U0 2 U0 2 9.Microstructures of U 3 SiAl (U + 3.5 wt% Si + 1.5 wt% Al). Curves A-D show heat-flow data for fuels with x -1.58, 1.65, 1.7, and 1.8, respectively. Lines UL, LL, and MP indicate the upper limit, lower limit, and most probable value, respectively, for the thermal conductivity of fuel with x -1.6. Line A shows pulse-heating data for fuel with x -1.6, which do not agree with curve MP. ABSTRACTA thorough review is provided on nuclear fuels for steady-state thermal research and test reactors.The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current U.S. policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on U.S. activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy and the UAl -Al and U 3 0 8 -Al dispersion plate fuels, and the U-Al alloy, UZrHx, and U0 2 rod fuels. Among new fuels, those given major emphasis include U 3 Si-Al dispersion and b0 2 caramel plate fuels.

Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100dpa

“…The development of spot features at high doses (>80 dpa) indicates that the microstructure evolution continues beyond 100 dpa at this irradiation temperature. Considering the melting temperature of $1350°C for UAl 3 type phase [14], the development of large loops at a relatively low irradiation temperature (T irr /T melt = 0.29) suggests a high mobility of point defects in this phase. Superdislocations can form in the ordered structure, such as intermetallic compounds of AB 3 or A(B,C) 3 -type fcc structures where atom-A occupies the corner sites of the cubic unit cell [15][16][17].…”

Review of oxidation rates of DOE spent nuclear fuel : Part 1 : nuclear fuel.

“…There were a few studies that investigated the corrosion behavior of Al-U 3 O 8 dispersion fuels with defected cladding. 127,128,129 Peacock recently reviewed the properties of Al-U 3 O 8 fuel. 130 Peacock considered the corrosion behavior generally the same as that of the aluminum matrix, since the reaction rate of U 3 O 8 with water is very small in comparison.…”

Dispersion Fuels

“…Since there is no apparent reason for the UA1 X -A1 thermal conductivity to be significantly lower than that for U 3 O 8 -A1 fuel, one suspects that the data may be in error. In fact, Stahl (1982), in an earlier compilation of thermal conductivity data, concluded that these particular specimens were not well-characterized and that the measurement system was not well-calibrated. One might expect UAl^ dispersions to behave like U 3 Si 2 dispersions since they have many other similarities.…”

“…It has generally been assumed (Stahl, 1982) that the thermal expansion of the dispersion is essentially controlled by the expansion of the Al matrix, since the expansion coefficient for Al is 1.5 to 3 times that of the uranium aluminides (~ 16 x 10" 6 ), the uranium oxides (~ 9 x 10 " 6 ), and the uranium silicides (~ 15 x 10 ~6). However, measurements performed by Nazare et al (1975) showed that this assumption is not completely valid.…”

“…(1966), Posey (1983), Shibata et al (1984), Woodley (1986Woodley ( ,1987, and Saito et al (1989) have performed measurements of fission product release from aluminum-based dispersion fuels. Stahl (1982) provided an excellent summary of the information available before 1982, both from out-of-reactor experiments and from reactor accidents. Taleyarkhan (1990) performed a detailed statistical analysis of most of the available data.…”