Implementation of zirconium-based components in a nuclear reactor core, in fuel cladding or advanced composite fuel coatings and matrix, is enabled by its low thermal neutron capture cross-section. Low — activation structural materials are desired, and par­ticularly for ZrC in fuels, diversion of neutrons from fission reactions is to be avoided. The isotopic average cross-section for Zr for 0.0253 eV thermal neutrons (at room temperature) is 0.19 barn (1.9 x 10- 9m2), and for C is 0.0035 barn (3.5 x 10-31 m2).160 Owing to similar chemistry, Zr naturally occurs together with up to 2 wt% Hf unless specially purified for nuclear reactor applications (<100 ppm Hf). The isotopic average cross-section for Hf is 106 barn (1.06 x 10-26 m2), making it a very undesirable impu­rity in Zr for reactor applications.

2.13.6.2.1 Durability and dimensional stability under neutron irradiation

Keilholtz eta/.161 irradiated hot-pressed, slip-cast, and explosion-pressed ZrC at 403-628 K (temperatures corrected by Watson and Keilholtz162) with a fast neutron (>1 MeV) fluence of (0.4-5.4) x 1021 cm-2, including 63-142 thermal cycles from room temper­ature to the irradiation temperature during the experiment, in the Oak Ridge National Laboratory’s Engineering Test Reactor. Damage due to thermal cycling alone was ruled out by out-of-pile tests. Severe fracturing of the hot-pressed and slip-cast samples were noted for fluences above (2.5-3) x 1021cm-2, with the explosion-pressed (containing Co or Ni binder) samples showing minor to severe damage above 1.1 x 1021cm — . Of the 2-3% radiation-induced volumetric swelling measured for all sample forms, 1% was accounted for by lattice parameter expansion, with the authors proposing point defect clusters and gas bubbles as causing the remainder. Helium gas is produced in carbides through fast-neutron reactions with carbon, and the authors considered the possibility of hydrogen pro­duced by thermal neutron reactions with nitrogen impurities. Swelling increased with fluence up to 2 x 1021 cm — , but further fluence resulted in the same or lesser degree of expansion.

Keilholtz et a/.163 expanded this study to irradiation at 1273-1373 Kat afluence of2.4 x 1021 cm-2. Swelling was lower at the higher temperatures, with <1% volume expansion measured. Watson and Keilholtz162 irradiated the same materials at low temperature and fluence (338-373 K, 0.21-0.72 x 1021cm-2), with

1. 7-2.5% volume expansion reported. The irra­diated samples were annealed at 973-1373 K until dimensional change ceased. Volume contracted with increasing temperature up to about 1173 K, with the degree of contraction saturating at higher tempera­ture. Some evidence for reexpansion of slip-cast ZrC at higher temperatures was reported, but there were too few data points to determine this conclusively. The authors proposed that expansion/contraction behavior as a function of fluence was caused by atomic displacements and point defect clusters produced by initial fast neutron cascades, with subsequent cascades at higher fluence returning some displaced atoms to lattice sites. Contraction was also attributed to the annealing of defects dur­ing postirradiation heat treatment. As temperature increases, easily annealed single point defects are consumed, leaving more difficult-to-anneal defect clusters, and further contraction is limited.

These observations are consistent with the find­ings of Andrievskii et a/.,164 who irradiated sintered ZrCo.98 at either 423 or 1373 K with a fast neutron fluence of 1.5 x 1020 cm—2 Lattice parameter expanded with irradiation but to a lesser degree at the higher temperature: 0.47% at 423 K versus 0.13% at 1373 K (volume increase of 1.4% vs. 0.38%). Correspond­ingly, density decreased after irradiation; however, the decrease was more, not less, pronounced at a higher temperature (—1.7% at 423 K vs. —2% at 1373 K). The authors propose that despite a lesser degree of lattice parameter swelling at higher tem­peratures, intergranular porosity increased and accel­erated swelling.

Koval’chenko and Rogovoi165 irradiated ZrC0 98 at 323 K with thermal neutrons to a fluence of 1 x 1019—1.5 x 1020cm— , observing a lattice param­eter expansion of 0.07-0.71% (volume expansion

0. 21-2.15%) that increased with fluence. No change in the porosity or grain size of the samples was detected. Andrievskii et a/.164 attributed their lower lattice swelling to higher sample purity, as the sam­ples of Koval’chenko and Rogovoi165 had low initial lattice parameter and were likely high in O and N impurities.

Andrievskii et a/.166 considered swelling as a func­tion of C/Zr ratio. They irradiated ZrC07-094 at 413 K in a fast neutron of fluence of 1 x 1019cm— . Lattice parameter increased after irradiation for all samples, but superimposed on the expected trend of increasing lattice parameter with increasing C/Zr ratio was a more pronounced increase for samples closer to stoichiometry. Lattice parameter of ZrC073 expanded by 0.04% (0.12% volume expan­sion), while ZrC094 expanded by 0.33% (0.99% volume expansion). The authors proposed that in near-stoichiometric compositions, fast neutron-induced atomic displacements may force C atoms into tetrahe­dral interstitial sites, while in more nonstoichiometric compositions, C atoms may be displaced to already vacant octahedral sites. In this sense, the authors judged irradiation-induced nonstoichiometry in more carbon — deficient ZrCx to be qualitatively similar to unirradi­ated nonstoichiometric ZrCx.

ZrC has also been subjected to high-temperature heat treatment and irradiation as a 20-50 q. m thick coating layer in testing of TRISO fuel having a cen­tral actinide fuel kernel coated in a porous carbon buffer layer, a dense pyrocarbon layer, a ZrC layer, and an outer pyrocarbon layer. Particles have been irradiated loose or embedded in the matrix of a fuel compact. The performance of ZrC in this configuration is inevitably tied to that of the system as a whole, as the neighboring coatings and fuel matrix not only add structural support but also intro­duce the possibility of reactions among ZrC, fission products, gases, graphite, and actinides.

Reynolds eta/.167 fabricated ZrC coatings either in the standard TRISO configuration or directly on top of fissile 235UC2 or fertile (232Th,233U)O2 kernels, the latter designed as a test of ZrC corrosion. These were irradiated at 1473 K with a fast neutron (>0.18 MeV) fluence of 5 x 1021 cm—2, to a burnup of 70% fissions per initial actinide metal atom (%FIMA) for fissile and 8% FIMA for fertile particles, in the Commissar­iat a l’Energie Atomique’s Siloe Test Reactor. Postir­radiation examination revealed that the only ZrC damage was radial cracking in almost all the particles with ZrC directly atop kernels; as this configuration lacked buffer volume to accommodate fission gases, such failures were expected. ZrC in the standard TRISO configuration was crack-free. No diffusion or reactions between actinides/fission products and ZrC was evident in any configuration.

Ogawa et a/.168 fabricated TRISO particles with a 20-30 qm thick ZrC layer and UO2 kernel and compacted these in a graphite-resin matrix at 2073 K. Irradiation by fast neutrons (>0.18 MeV) under various conditions (1173-1873 K, fluence (1-2.2) x 1021 cm—2, burnup of 1.5-4% FIMA, during 81-156 effective full-power days) was carried out in the Japan Materials Test Reactor at the Japan Atomic Energy Research Institute. Following decon­solidation of the fuel compacts to recover loose par­ticles, ceramography showed no ZrC degradation in particles irradiated at the highest burnup, fluence, and temperature for over 135 effective full-power days. A zero particle failure rate was assessed by in-reactor monitoring of fission gas 85Kr release rate during 80 effective days of irradiation at 1173 K, which was an order of magnitude smaller than that expected for a single particle rupture out of 7000 particles in this test.

Minato eta/.169 irradiated similarly prepared com­pacts between 1673-1923 K with a fast neutron (>0.18MeV) fluence of 1.2 x 1021cm—2 to 4.5% FIMA burnup during 100 effective full-power days, finding a through-coating failure rate of 0.01%, that is, failure of less than one out of 2400 particles irradiated.

Some of the particles irradiated by Ogawa eta/.168 were subsequently heat treated at 1173-2273 K. Par­ticles originally irradiated at 1173 K to a fluence of

1.2 x 1021 cm—2 and burnup of 1.5% FIMA during 80 effective days were annealed in flowing He by Minato et a/.170 at 1873 K for a total of 4500 h and by Minato et a/.120,171 at 2073 K for 3000h and at 2273 K for 100 h. In situ 85Kr release monitoring showed none of the 100 particles annealed during each test ruptured. Ceramography and X-ray micro­radiography ofparticles from each test showed that at 1873 K, there was no thermal degradation or corro­sion of ZrC coatings. At 2073 K, the ZrC coating was intact, but there was some surface roughness attrib­uted to thermal degradation and evidence of attack along the grain boundaries. At 2273 K, all but 7 of the 100 particles showed some failed or damaged coating layers, including the ZrC layer, and evidence of reac­tion or interdiffusion between ZrC and U.

Interpreting the results via a thermodynamic analysis of the Zr-C-U-0 system, the authors attri­bute deterioration of ZrC at these temperatures, at least in part, to mechanical failure of the inner pyrocarbon layer combined with oxidation of carbon by the oxygen released during the transmutation of U in U02. Subsequent exposure of ZrC to C0(g) oxidizes ZrC to Zr02 and C and reduces the ZrC coating integrity.

Durability of ZrC is more likely to be limited by irradiation of the system contained within rather than by high-temperature degradation of ZrC alone. Ogawa and Ikawa172 subjected unirradiated TRISO particles having either U02 or (Th, U)02 kernels to annealing in He atmosphere for 1h at 2073 K (as during fuel compact fabrication) followed by 1 h at 2173-2823 K. ZrC coatings on all (Th, U)02 particles were intact after 1 h at 2823 K, but durability of U02 particles was guaranteed only to 2373 K, with swelling noted at 2723 K and U migration out of the ZrC coating at 2773 K. ZrC grain growth and plastic deformation occurred, especially above 2723 K.