I.J. Hastings

Temperature-dependent swelling in irradiated U3Si fuel elements

Abstract The volume changes in irradiated U3Si are shown to be strongly temperature-dependent. Density values were determined as a function of radius for a series of fuel elements which had operated at calculated centre temperatures between 475 and 675 °C. The fuel surface temperatures ranged from 315 to 440 °C, and the burnups from 1500 to 8400 MWd/tonne U. The volume change, †V/V , increases from about 5% at 315 °C to a maximum of about 30% at 500 °C, then decreases to †V/V on burnup was noted in the range examined. An axial void of adequate dimensions can accommodate volume changes if the central temperature of the fuel is in the range where maximum swelling occurs. If the largest volume increase occurs close to the fuel surface, the axial void is less effective, and a diametral increase to accommodate the swelling may result.

Release of short-lived fission products from UO2 fuel: Effects of operating conditions

Abstract To support accident “source-term” studies, we have carried out a series of in-reactor tests to determine the behaviour of short-lived fission products (Xe, Kr, I) in operating UO 2 fuel under a variety of conditions. The linear power range examined was 1762 kW/m to a maximum burnup of about 220 MW. h/kg U. Under normal operating conditions, release of the short-lived xenons and kryptons was described by a λ(decay constant) −0.5 relationship, characteristic of diffusion; there was a strong temperature-dependence of release in the thermally-activated diffusion range. Release during shutdown and startup transients was dependent on prior operating temperature, while under oxidizing conditions enhancement of release up to a factor of 4.5 was observed. Under loss-of-coolant conditions, release was described by a λ −1 relationship, characteristic of movement of stored inventory. Most release accompanied the re-wet portion of the transient, consistent with enhancement due to fuel cracking. No iodines exited from the fuel elements under normal, oxidizing or loss-of-coolant conditions.

Bubble formation in irradiated Li2O

Abstract Lithium oxide, irradiated to a burnup of 1 at% (total lithium) at temperatures between 400 and 850°C with on-line tritium recovery and measurement, has been examined out-reactor. Residual tritium content ranged from 2.4 to 16 mCi/g, but, conservatively, averaged less than 10 mCi/g or 1 wppm. Scanning electron microscopy showed bubble formation in the ceramic which is thought to be due to helium formed from the in-reactor 6Li(n, α)3H reaction.

BEATRIX-II: A multinational solid breeder materials experiment

BEATRIX-II is an in situ tritium recovery experiment in the fast flux test facility (FFTF) reactor designed to characterize the feasibility of utilizing solid breeder materials at extended burnups in a fast neutron flux. Although not yet complete, the BEATRIX-II experiments have already substantiated that the solid breeder selected for ITER, Li 2 O, has good irradiation stability and tritium recovery. Temperature stability, lithium transport, dimensional stability and tritium recovery issues of Li 2 O up to 5% Li burnup were addressed in this experiment, Temperature gradients far more severe than in the ITER design, 400 to 1000°C, were found to be essentially unchanged by burnup and produced no observable instability, either from swelling or lithium vapor transport. Temperature change experiments illustrated that lithium inventories do not appear to increase as a result of irradiation to burnups of 5%.

Fission gas release from power-ramped UO2 fuel

Abstract Power-ramped UO 2 fuel released up to 14% fission gas, depending on power history; equivalent unramped fuel released less than 1%. Fuel in ramped elements showed grain growth, intergranular fission gas bubbles, grain-edge tunnels and grain boundary separation (microcracking). The significance of release during the post-transient temperature decrease should be examined further.

Burnup and temperature dependence of swelling in U3Si

Abstract The initial swelling in irradiated U 3 Si fuel elements is dependent on burnup and temperature. The volume increase, ΔV / V , was determined as a function of fuel radius in elements which operated at calculated central temperatures between 530 and 600 °C, and surface temperatures from 350 to 380 °C. Fuel burnups varied from 50 to 990 MWd/tonne U. There was no significant swelling up to at least 270 MWd/ tonne U, with the first such indication occurring at 795 MWd/tonne U, where the peak ΔV / V value was 8%. At 990 MWd/tonne U, the maximum volume increase was 11.5%. At both burnups the swelling was temperature dependent, with the maxima in the range from 500 to 540 °C. Transmission electron microscopy of low burnup samples revealed defects which may represent an early stage in the growth of pores responsible for the swelling observed at higher burnups. Previous studies in the range 1500–8400 MWd/ tonne U have been extended to 14 700 MWd/tonne U; no further volume increase occurred with the additional burnup.

High burnup performance of annular UO2 fuel with inter-pellet graphite discs

Abstract We have irradiated annular UO 2 fuel with inter-pellet graphite discs at linear powers of 62 and 44 kW/m to a maximum burnup of 775 MWh/kgU (32000 MWd/TeU). The combination reduced fission gas release by up to a factor of four compared with that for annular fuel alone, for absolute releases up to 45%. Disc compatibility with other fuel components was acceptable.

Simulation of in-reactor swelling in U-3.5 wt % Si-1.5 wt % Al by ion bombardment

Ion bombardment of U-3.5 wt. % Si-1.5 wt. % Al has produced swelling which corresponds qualitatively with that observed in-reactor. Samples were bombarded with 0.5–2 MeV argon ions to damage levels corresponding with those produced by fluences of 2.8–8.1 × 1025 fissions/m3; temperatures were 570–950 K. Up to 200 vol % swelling was recorded in unrestrained samples, with a maximum at 670–870 K. The results are interpreted in terms of swelling due to void/bubble formation.

Canadian fusion breeder blanket program: Irradiation facilities at chalk river

Abstract The major irradiation facility at Chalk River Nuclear Laboratories (CRNL) is the NRU research reactor. Both unvented and vented capsule experiments on candidate blanket ceramics can be performed. In the unvented tests, tritium release data (HT-to-HTO ratio, tritium retention) are obtained by post-irradiation heating of the breeder ceramic in the presence of a sweep gas. Four tests have been completed on Li 2 O and LiAlO 2 . Effects of sweep gas composition, extraction vessel material and ceramic properties have been determined. Two unvented irradiations under the BEATRIX international breeder exchange program have been completed; analysis is underway. The vented tests involve long-term irradiation of candidate blanket materials. CRITIC-I, scheduled for mid-1986 under BEATRIX, will examine ANL-fabricated Li 2 O in a six-month irradiation at 700–1200 K, varying sweep gas composition, with on-line HT/HTO measurement. Additionally, accelerator simulation techniques are available, using 70 kV and 2.0 MV mass separators, a 2.5 MV Van de Graaff accelerator and a tandem accelerator super-conducting cyclotron, the latter allowing irradiation with protons, deuterons or helium at 18–20 MeV.

Irradiation of U3Si-based compounds in the high-voltage electron microscope

U3Si and U-3.5 wt% Si-1.5 wt% Al have been irradiated in a high-voltage electron microscope (HVEM) at 500–1180 keV and 300–660 K. The creation of ‘black spot’ damage and removal of deformation twins are observed. Defects about 10 nm diameter averaging 4 × 1021 m−3 are seen only in samples pre-injected with 10−5 atomic fraction argon and are tentatively identified as voids. The atomic displacement rate during HVEM irradiation of U3Si-based compounds is about two orders of magnitude higher than that for fuel at power reactor ratings. It is inferred that displacement of silicon, and possible uranium, atoms in U3Si-based compounds occurred in the HVEM at accelerating voltages in the range 700–1180keV.