N.H. Miller

Oxidation of UO2 fuel by the products of gamma radiolysis of water

Abstract The kinetics of UO 2 fuel oxidation by the products of gamma radiolysis of water were studied as a function of absorbed dose rate. The oxidation of UO 2 during radiolysis was monitored by recording the corrosion potential of a UO 2 electrode as a function of time. Changes in the surface of UO 2 were determined using cathodic-stripping voltammetry (CSV) and X-ray photoelectron spectroscopy (XPS). The effects of specific radicals on UO 2 oxidation were determined by adding suitable scavengers to the solution to maximize the yield of a particular radical. The radical species formed during radiolysis of water are much more effective in causing UO 2 oxidation than the molecular oxidants formed during radiolysis of water or present in water from atmospheric sources. The oxidation of UO 2 in the presence of oxidants produced by the gamma radiolysis of water occurs in two distinct stages: (a) the formation of a thin layer ofUO 2 + x with a stoichiometry close to UO 2.33 and a thickness similar to that obtained (over longer exposure periods) in unirradiated oxygenated solutions; and (b) the subsequent oxidative dissolution of this surface layer to produce soluble U VI species and secondary phases, probably hydrated schoepite (UO 3 · xH 2 O), on the UO 2 substrate. The first stage occurs in the potential range− 500 mV PCORR ≲−100 mV (versus SCE). The second stage starts around E CORR ∼− 100 mV and eventually achieves steady state at a value of E CORR determined by the gamma dose.

Oxidation of CANDU uranium oxide fuel by air in gamma radiation at 150°C

Oxidation of CANDU nuclear fuel (U02), heated at 150°C in air, in O2 with 60% saturated steam, and in Ar with 60% saturated steam atmospheres for ∼ 2 years in gamma fields of dose rate ∼ 15 Gy h−l, was investigated using X-ray photoelectron spectroscopy, X-ray diffraction and scanning electron microscopy. Surface analysis of U02 disks shows formation of U3O8 on the U02 disks exposed to air or O2. This is the first observation of the formation of U3O8 on UO2 by air oxidation at such a low temperature. The rate of U3O8 formation by dry air oxidation of UO2 at 150°C is very low. The presence of water vapor along with O2 increases the oxidation of UO2 in gamma fields, leading to the formation of U6+ phases, e.g., UO3 · χH2O along with U3O8. UO2 did not suffer any oxidation by water vapor radiolysis at 150°C in an O2-free atmosphere.

Corrosion of used nuclear fuel in aqueous perchlorate and carbonate solutions

Abstract The corrosion of used fuel was investigated using electrodes constructed from fuel pins discharged from the Pickering, Bruce and Darlington CANDU reactors, and compared to the corrosion behaviour observed on unirradiated UO 2 and SIMFUEL. Experiments were carried out in solutions of NaClO 4 (pH∼ 9.5) in the presence and absence of (a) substantial concentrations of sodium carbonate, and (b) additional external gamma fields. Used fuel electrodes reached oxidizing corrosion potentials ( E CORR ) rapidly compared with unirradiated UO 2 electrodes. However, optical and SEM examinations showed no evidence for rapid oxidative dissolution. This reaction, expected to be fast since high values of E CORR are observed, appears to be blocked by the accumulation of secondary phases in grain boundaries. The oxidation and dissolution behaviour of used fuel is determined predominantly by (i) the dose rate in solution near the fuel surface, (ii) the extent of burnup (which determines the degree of fission product doping), and (iii) the degree of non-stoichiometry.

Electrochemical studies of SIMFUELS

Abstract The properties of a standard UO 2 fuel and three SIMFUEL materials (unirradiated analogues of used nuclear fuels) have been characterized electrochemically and under natural corrosion conditions. The general oxidation and reduction behaviour was determined voltammetrically. The effect of substituents on the surface reactivity was determined by studying the electroreduction of O 2 . Under natural corrosion conditions, SIMFUELS oxidize more rapidly than does UO 2 , but then appear to dissolve at a similar rate.

XPS and XRD studies of samples from the natural fission reactors in the Oklo uranium deposits

Mineral samples from the natural fission reactors 10 and 13 in the Oklo uranium deposits were studied using X-ray photoelectron spectroscopy (XPS) and X-ray diffraction (XRD) to gain information about the long-term behaviour of UO{sub 2} fuel in a geological disposal vault. Two samples from reactor zone 10 (samples No. D81N-190292 and D73-88) and one sample from reactor zone 13 (sample No. SD37-S2/CD) were analysed. Low-resolution XPS spectra were recorded to determine the major elements present in the ore. High-resolution spectra were recorded to gain information about the chemical state of the elements present in the mineral samples. The samples show low values for the U{sup 6+}/U{sup 4+} ratio. The oxidation state of uranium in these samples is even lower than that in U{sub 4}O{sub 9}.The binding energies of the Pb 4f bands indicate most of the Pb is in the +2 oxidation state in these samples. The C ls band indicates the presence of organic carbon. XRD analysis shows that the main uranium-bearing phase is uraninite and lead is present mainly as galena. The significance of the results for nuclear fuel waste management is discussed.

Prediction of the Oxidative Dissolution Rates of Used Nuclear Fuel in a Geological Disposal Vault Due to the Alpha Radiolysis of Water

Effects of alpha radiolysis of water on the corrosion of nuclear fuel (UO 2 ) have been investigated in solutions at pH = 9.5, i.e., a value close to that expected in groundwaters at the depth of the disposal vault proposed in the Canadian nuclear fuel waste management program, CNFWMP. The corrosion potentials of UO 2 electrodes exposed to the products of alpha radiolysis of water were monitored as a function of alpha flux and exposure time in a specially designed thin-layer cell. The oxidative dissolution rates of UO 2 are calculated from the steady-state values of the corrosion potential using an electrochemical model. A procedure to predict the dissolution rate of used nuclear fuel in groundwater as a function of fuel cooling time is described, and illustrated by calculating the dissolution rates for the reference used fuel in the CNFWMP (Bruce CANDU reactor fuel, burnup 685 GJ/kg U). It is shown that the oxidative dissolution of used fuel in the CNFWMP will be important only for time periods ≤ 600 a at this burnup and assuming no decrease in pH.