Uranium-series disequilibria data, in conjunction with petrographic analyses, indicate that the uranyl oxide hydrate becquerelite can persist for hundreds of thousands of years, possibly longer. Becquerelite probably forms continuously as ground water compositions permit and is resistant to U leaching by ground water. On the time scale of interest for the geologic disposal of spent UO2 nuclear fuel, becquerelite is a long-lived sink for uranium in oxidizing, U and Ca-bearing ground waters. Such long-term stability also supports recent solubility experiments that indicate natural becquerelite has a lower solubility product than that determined for synthetic becquerelites.

The accurate identification of uranium-bearing corrosion products from spent fuel leaching studies is often difficult because the XRD powder patterns for many U(VI) phases are closely similar. Standard pattern-matching techniques commonly used for phase identification (e.g., JCPDS) make confident phase identification difficult and can oversimplify complex mixtures of UOHs. Refinement methods that fit the full XRD powder pattern, such as Rietveld, may be the only methods for accurate UOH identification, especially for data from sample mixtures. Rietveld refinement of XRD powder data for uranium oxides can be accomplished for phases whose structures are well defined. Unfortunately, many of the uranyl oxide hydrates are still ill-defined, both structurally and compositionally. The uranyl oxide hydrates have large unit cells and relatively low symmetries, as well as close structural similarities, all of which complicates the simultaneous refinement of mixed UOH phases. Modeling peak shape profiles can also hamper refinements of mixed-UOH samples due to extreme peak overlap. Because of these difficulties, Rietveld structure determination for UOHs from sample mixtures remains intractable, although structure determination may be possible for fine-grained, single-phase samples. Nonetheless, Rietveld refinement is the most promising method for accurately identifying the component phases in a mixture, but the unknown UOH structures must be determined before this technique can be applied successfully.

Localized corrosion in aqueous environments forms an important bounding condition for the performance assessment of high-level waste (HLW) container materials. A predictive methodology using repassivation potential is examined in this paper. It is shown, based on long-term (continuing for over 11 months) testing of alloy 825, that repassivation potential of deep pits or crevices is a conservative and robust parameter for the prediction of localized corrosion. In contrast, initiation potentials measured by short-term tests are non-conservative and highly sensitive to several surface and environmental factors. Corrosion data from various field tests and plant equipment performance are analyzed in terms of the applicability of repassivation potential. The applicability of repassivation potential for predicting the occurrence of stress corrosion cracking (SCC) and intergranular corrosion in chloride containing environments is also examined.

This report focuses on the prediction of materials performance for the carbon steel corrosion-allowance container as a function of thermal loading for the potential repository at Yucca Mountain. Low, intermediate and high thermal loads were evaluated as to their performance given assumptions regarding the temperature and humidity changes with time and the resultant depth of corrosion penetration. The reference case involved a kinetic relation for corrosion that was utilized in a sensitivity analysis to examine the impacts of time exponent, pitting, and microbiologically-influenced corrosion. As a result of this study, the high thermal load appears to offer the best corrosion performance. However, other factors must be considered in making the final thermal loading decision.

A physically-based, phenomenological stochastic model for pit initiation and growth is presented as a potential tool for predicting the degradation of high-level radioactive-waste containers by aqueous pitting corrosion. Included in the model are simple phenomenological equations describing the dependence of the controlling stochastic parameters on the applied (or corrosion) potential, chloride ion concentration, and absolute temperature. Results from this model are presented that demonstrate its ability to simulate several important environmental effects on pitting.

In the Swedish waste management programme, the copper canister is expected to provide containment of the radionuclides for a very long time, perhaps millions of years. The purpose of the present paper, is to analyse prerequisites for assessments of corrosion lifetimes for copper canisters.

The analysis is based on compilations of literature from the following areas: chemical literature on copper and copper corrosion, mineralogical literature with emphasis on the stability of copper in near surface environments, and chemical and mineralogical literature with emphasis on the stabilities and thermodynamics of species and phases that may exist in a repository environment.

Three main types of situations are identified: (1) under oxidizing and low chloride conditions, passivating oxide type of layers may form on the copper surface; (2) under oxidizing and high chloride conditions, the species formed may all be dissolved; and (3) under reducing conditions, non-passivating sulfide type layers may form on the copper surface.

Considerable variability and uncertainty exists regarding the chemical environment for the canister, especially in certain scenarios. Thus, the mechanisms for corrosion can be expected to differ greatly for different situations. The lifetime of a thick-walled copper canister subjected to general corrosion appears to be long for most reasonable chemistries. (It is assumed that the canister has no defects from manufacturing and that the bentonite buffer is intact). Localized corrosion may appear for types (1) and (3) above but the mechanisms are widely different in character. The penetration caused by localized corrosion can be expected to be very sensitive to details in the chemistry.

The finite-difference CAMEO code for modeling general corrosion of copper canisters is described. CAMEO represents the engineered barrier system and surrounding fractured host rock in 3-dimensional cylindrical coordinates. The time of containment failure is evaluated using CAMEO, as constrained by transport rates of corrodants to the canister or by transport rates of corrodant products away from the canister. Additional chemical processes explicitly modeled in CAMEO include 1) copper corrosion, and 2) kinetics of Cu(I) oxidation to Cu(II), both as a function of near-field pore water chemistry, specifically pH, Eh, and chloride concentration. The diffusional transport and sorption behavior of Cu(I) and Cu(II) are also separately incorporated.

This paper describes a joint modelling and experimental study for investigation of pit growth in carbon steel High-Level Radioactive Waste overpacks under consideration in Japan. A mathematical model of the growth of corrosion pits in metals has been developed. This model is implemented in the computer program CAMLE, and includes representation of the chemical, electrochemical and migration processes that control pit-growth rates. Experiments to provide key input data for the model are described, in addition to experiments to measure ‘short-term’ pit growth. Predictions from the model are compared with these data. Overall, the comparisons are encouraging and the model shows good potential as a tool for assessment of the long-term corrosion behaviour of overpacks under repository conditions. Future developments of the model to improve agreement are discussed.

The use of bentonite as buffer and carbon steel as overpack material for the geological disposal of nuclear waste is under investigation. To better assess the long term integrity of the carbon steel overpack, a quantitative analysis of the corrosion behavior on the steel surface for time frames beyond that of feasible empirical determination is required. The state n years after disposal, consisting of Carbon Steel / Corrosion Products + Bentonite / Water, was simulated and the corrosion behavior of the carbon steel in this state investigated. The following facts became apparent. Both the corrosion rate and the non-uniformity of it increased with increase in the corrosion product content in the compacted bentonite. When the corrosion product layer was formed between the carbon steel and the bentonite, it ennobled the corrosion potential and increased the corrosion rate.

As a sequence of studies to evaluate the quantity of gas evolution from low/intermediate level waste repositories,hydrogen gas evoluted from corrosion of carbon steel in simulated repository environment was evaluated by laboratory experiments. The experimental results on the hydrogen gas evolution both in air purging condition simulated oxidizing environment and nitrogen purging condition simulated reducing environment, are summarized as follows.

1. (1) Hydrogen gas evolution enough to analyze quantitavely by gas chromatography (>5ppm) has been recognized under almost all test conditions except reducing equilibrium cement water.

2. (2) Effects of purging gas (air,nitrogen) on the hydrogen gas evolution and the corrosion rate calculated from weight loss were air purge > nitrogen purge. On the other hand, the contribution ratio of hydrogen evolution reaction in corrosion rate was nitrogen purge > air purge.

3. (3) Effects of test solution on the hydrogen evolution rate were as fo11ows.

   • • Air Purge : Equilibrium Bentonite Water ≈ Equilibrium Cement Water > Synthetic Sea Watert

   • • N2 Purge: Synthetic Sea Water > Equilibrium Bentonite Water >> Equilibrium Cement Water

4. (4) No distinct effect of crevice geometry of test specimen on hydrogen evolution rate was recognized. Only under the reducing equilibrium cement water, however, the increase of hydrogen evolution was confirmed after the immersion of several hundred hours.

5. (5) Hydrogen evolution rates tended to decrease with testing time except in the reducing equilibrium cement water.

6. (6) No distinct difference of hydrogen evolution rate between steels (SPHC, SPCC) was observed.

The repassivation potentials, Er.crev’s, for metal/metal-crevice of Alloy 625 were determined in 0.3–10% NaCl solutions for temperatures up to 250 C. The Er.crev’s were found to be the least noble at temperatures around 100 and 125 C. The Er.crev became more noble as temperature increased; this tendency was particularly strong in diluted solutions. Based on the experimental data, a crevice corrosion map showing the critical condition in terms of temperature and NaCl concentration was presented. As for the map, a wide repassivation region was found in elevated temperatures, similar to that of commercially pure titanium, C.P.Ti.

The photoelectrochemical behaviours of a TiO2 single crystal and TiO2 coating were studied, for the purposes of cathodic protection of stainless steels and Cu via the TiO2 coating combined with glass scintillators under gamma ray irradiation. It was confirmed that a TiO2 coating could protect 304 stainless steel cathodically from crevice corrosion under illumination. A logarithmic relationship between the photopotential of single crystal TiO2(rutile) and light intensity was found, moreover, the photopotential was found to be least noble when wavelength equals 375 nm. Under illumination by gamma rays combined with the glass scintillators, the electrode potential of single crystal TiO2 was found to shift in the less noble direction by about 200 mV. Therefore, the technique of cathodic protection by TiO2 coating is considered to be applicable to protect the packaging metal from corrosion for a long time.

SKB has been evaluating a copper/steel canister for use in the disposal of spent nuclear reactor fuel. The canister consists of an outer layer of about 50 mm thickness of copper to provide corrosion resistance and an inner layer of similar size to provide structural strength. Once the canister is breached by corrosion, it is possible that the void volume inside the canister might fill with water. Water inside the canister would moderate the energy of the neutrons emitted by spontaneous fission in the fuel. If the space in the canister between and around the fuel pins is occupied by canister filling materials, the potential for criticality is avoided. We have developed a set of design requirements for canister filling material for the case where it is to be used alone, with no credit for burnup of the fuel or other measures, such as the use of neutron absorbers. Requirements were divided into three classes: essential requirements, desirable features, and undesirable features. The essential requirements are that the material fill at least 60% of the original void space, that the solubility of the filling material be less than 100 mg/l in pure water or expected repository waters at 50°C, and that the material not compact under its own weight by more than 10%. In this paper we review the reasons for these requirements, the desirable and undesirable features, and evaluate 11 candidate materials with respect to the design requirements and features. The candidate materials are glass beads, lead shot, copper spheres, sand, olivine, hematite, magnetite, crushed rock, bentonite, other clays, and concrete. Emphasis is placed on the determination of whether further work is needed to eliminate uncertainties in the evaluation of the ability of a particular filling material to be successfully used under actual conditions, and on the ability to predict the long-term performance of the material under the repository conditions.

New X-ray diffraction and scanning electron microscopy data are given for the incorporation of Np and Pu in zirconolite, at levels of tens of percent. The actinide valences and the cations they replace are deduced from the microanalysis of the zirconolite compositions, and X-ray absorption data are used to obtain more direct information on the valences of Ce and Nd, which are used as simulants of Pu and trivalent actinides respectively. Trivalent rare earths and actinides have extensive solid solubility in zirconolite, mainly but not exclusively in the Ca site. Tetravalent rare earths and actinides have considerable solid solubility in the Zr site of zirconolite, and some solubility in the Ca site, but the strong tendency of zirconolite with ions substituted in the Zr site to undergo phase separation complicates structural interpretation. In zirconolite-rich Synroc-type ceramics designed to immobilise waste actinides, the target actinide waste loading has been set at 20 wt% and early leach results indicate the durability is at least as good as that of Synroc-C.

The immobilisation of excess weapons plutonium into Synroc can meet all of the important criteria discussed by the U.S. National Academy of Sciences (NAS)(1) for disposal, if the disposal option is pursued rather than options that exploit the energy value of plutonium. This paper summarises the relevant background of Pu incorporation into Synroc, the durability of Pu-containing Synroc and outlines a process flowsheet based on the experience with the 10 kg/hr Synroc Demonstration Plant at ANSTO. The extensive solid solubility of Pu in Synroc, coupled with a very high degree of chemical durability under hydrothermal conditions, makes Synroc ideal as a waste matrix for Pu disposal in deep boreholes to minimise diversion and proliferation risks.

The local structures around Zr atoms such as coordination number of oxygen and cations, and the Zr-0 and Zr-Ln (Ln; La, Ce) interatomic distances in the (La1-xCex).2Zr2O7 Pyrochlore compounds (X = 0–0.25), where Ce is a simulation of actinides,were studied by means of extended X-ray absorption fine structure (EXAFS) spectrometry. The local structure determined by EXAFS in this study is discussed in relation to the dissolution behavior of (La1-xCex)2Zr207 (leaching rates of Zr, La and Ce) in an acid solution (pH=1) at 363K reported previously. The difference in the local structure before and after leaching test was also investigated.

Hydroxyapatite (HAP : Ca10(PO4 ) 6 (OH) 2 ) and aluminosili-cate were easily synthesized using a hydrothermal process with mixing ratios of Ca/P=1.67 (Ca(PO3)2 and Ca(OH)2) and Si/Al=4 (SiO2 and Al(OH)3). The temperature was kept constant at 573 K for 60 minutes. .Cesium (Cs) and Cerium (Ce) contained in Ca(PO3)2 , were found to take on a stable form in the pollucite (CsAlSi2O6) structure and the monazite (CePO4) structure.

Thorium and uranium phosphates were synthetized via soft chemistry and then calcined at high temperature, respectively up to 850 and 1170°C. Th3(PO4)4 was produced in agreement with published results.while the analog uranium phosphate : U3(PO4)4 was not found. Instead, we found we had made a new mixed valence compound : U(UO2)(PO4)2, for which crystal structure, absorption spectrum in ultraviolet and visible, and X-ray photoelectron spectroscopy were done. All of these experiments confirmed the presence of uranium in both tetravalent and hexavalent states.

Lixiviation of thorium phosphate labelled with radioactive isotopes : 227Th and 223Ra allowed us to determine the upper limit of the solubility for this compound = 7.6xl0–11 M and, the percentage of 223Ra released in deionized water (about 5xl0–3%).

Similar lixiviation tests on U(UO2)(PO4)2 labelled with 230U (radiochemically produced) gave an uranium solubility value 10 times greater than uranium measured by laser spectrofluorimetry. This discrepancy is attributed to the radiation chemistry inherent in the use of radioactive isotope at high specific activity. The lixiviated uranium fraction from U(UO2)(PO4)2 is seven to eight orders of magnitude higher than for thorium. Unexpected behaviour was observed for uranium solubility which increases by two orders of magnitude after 2000 hours of lixiviation. This phenomenon is not yet understood.

Previous studies of ceramic crystalline waste forms, e.g. Synroc, tailored ceramics, and supercalcine, have concentrated on phases which are major constituents in the formulations: zirconolite, pyrochlore, hollandite, perovskite and zircon. These phases usually occur as members of multi-phase assemblages which are required for the incorporation of the wide variety of radionuclide elements present in the waste and the non-radioactive components added during reprocessing and pretreatment. The crichtonite structure (AM21O38 and A2M19O36), based on crystallo-chemical considerations and natural compositional analogues, may effectively incorporate both fission products and actinides. The naturally occurring crichtonite structure types include Sr (crichtonite), Ca and REE (loveringite), Na (landauite), REE and U (davidite), K (mathiasite), Ba (lindsleyite), and Pb (senaite), which are classified based on the dominant, large cations occupying the A-site. The crystal structure contains three types of sites of distinct size, from very large, M0, intermediate (M1, M3, M4, and M5), to small (M2). Numerous coupled substitutions within these cation sites allow for charge balance. Synthesis experiments were completed on the Ba-, Sr-, Ca-, and K-member compositions at 3 GPa and 1,150 °C. Low pressure synthesis should be possible, as natural minerals mostly occur in low-P systems. Reaction products were characterized by powder x-ray diffraction, scanning electron microscopy and electron microprobe analysis. In addition to the crichtonite phases, rutile, spinel, perovskite and armalcolite were identified as well. The Crichtonite structure type is estimated to accommodate waste loading of up to 30 wt. % PW-4B waste.

The potentiality of hydrous TiIV-Si oxide (TS) and hydrous TiIV-Zr oxide (TZ) for the selective sorption and immobilization of radioactive Co2+, Eu3+, Th4+ and UO22+ ions has been studied experimentally: The distribution coefficients and sorption capacity for these ions at various conditions in solutions, the structure and the composition of the exchangers in relevant ionic forms at high temperatures, and the leaching rates from the calcined exchangers into water and 0.01M HNO3 solution at room temperature. These results were discussed together with the results previously obtained for Cs+ and St2+, in order to evaluate the performance of TS and TZ as solidification media for radioactive liquid waste. TS takes up these cations more than TZ does in acidic solutions due to the higher acidity of the exchange site. In TZ, transition metal ions show high affinity despite the small distribution coefficients for representative metal ions. The dissolution of Co2+, Eu3+, and Th4+ is very slow from both calcined exchangers due to the crystallization of titanates of these cations. The leaching rates of Cs+, St2+ and UO22+ from calcined TS are much slower than those from TZ, indicating the formation of titanium silicate containing these cations. Based on these results, TS is considered to be superior to TZ in the sorption capacity and fixation capability of various kinds of radionuclides and is considered as a promising material for the solidification of radwaste.