This study aims at identifying the release mechanisms of helium in uranium dioxide. Two sets of polycrystalline UO2 sintered samples presenting different microstructures were implanted with 3He ions at concentrations in the region of 0.1 at.%. Changes in helium concentrations were monitored using two Nuclear Reaction Analysis (NRA) techniques based on the 3He(d,α)1H reaction. 3He release is measured in-situ during sample annealing at temperatures ranging between 700°C and 1000°C. Accurate helium depth profiles are generated after each annealing stage. Results that provide data for further understanding helium release mechanisms are discussed. It is found that helium diffusion appears to be enhanced above 900°C in the vicinity of grain boundaries possibly as a result of the presence of defects.

The presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable, leach resistant form. AWE has developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process using actinide-doped Type I waste (essentially a chloride-based waste) were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined using a modified MCC-1 test at 40°C/28 days were 12 x 10-6 g.m-2 and 2.7 x 10-3 g.m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (an oxyhydroxide-based waste) is currently underway at PNNL.

Differences in the ionic state of plutonium in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Early investigations into the use of HfO2 as the surrogate for Pu(IV) oxide in Type II waste showed some apparent differences in the phase assemblages of the surrogate and actinide-based products. However XRD examination of the products at higher resolution has demonstrated there is no significant difference and that for this work HfO2 is a suitable surrogate for PUO2.

Over extended periods, long-lived radionuclides (RN) or activation products within geologic disposal sites may be released from the fuel and migrate to the geo/biosphere. In the bedrock, contaminants will be transported along fractures by advection and retarded by sorption on mineral surfaces and by molecular diffusion into stagnant pore water in the matrix along a connected system of pores and micro-fissures.

The objective of this paper was to determine the connective porosity and mineral-specific porosities for three granite samples by 14C methylmethacrylate (14C-PMMA) autoradiography. Scanning electron microscopy and energy-dispersive X-ray analyses (FESEM/EDAX) were performed in order to study the pore apertures of porous regions in greater detail and to identify the corresponding minerals. Finally, the porosity results were used to evaluate the diffusion coefficients of RNs from previous experiments which determined apparent diffusion coefficients for the main minerals in three granite samples by the Rutherford Backscattering technique.

The total porosity of the Grimsel granite (0.75%) was significantly higher than the porosities of the El Berrocal and Los Ratones granites (0.3%). The porosities of the Grimsel granite feldspars were two to three times higher than the porosities of the El Berrocal and Los Ratones granites' feldspars. However, there was no significant difference between the porosities of the dark minerals. A clear difference was found between the various quartz grains. Quartz crystals were nonporous in the El Berrocal and Los Ratones granites when measured by the PMMA method, but the quartz crystals in the Grimsel granite showed 0.5% intra granular porosity. The apparent diffusion coefficients calculated for uranium diffusion within Grimsel granite on different minerals were very similar (210-13 ± 0.5 m2/s), but differences within both Spanish granites were found from one mineral to another (9 ± 110-14 m2/s in feldspars and 4.5 ± 0.510-14 m2/s in quartz) - always presenting lower diffusion values than in the Grimsel granite.

This study examined the conditions required to suppress microbial activity in compacted bentonite, such that microbially influenced corrosion (MIC) of copper waste containers, surrounded by compacted bentonite in a future deep geologic repository, would become insignificant. Experiments were carried out to determine the effects of dry density and porewater salinity on swelling pressure, water activity (aw) and the culturable microbial community in compacted bentonite. A dry density ³ 1.6 g/cm3 ensures that aw is < 0.96 and the swelling pressure > 2 MPa. Both conditions suppress microbial culturability below background levels (2.1 x 102 Colony-Forming Units/g) in as-purchased bentonite. Under such conditions, cells likely survive as dormant cells or inactive spores, which greatly reduces the possibility of significant MIC. Observations in natural clay-rich environments support these findings.

A mineralogical and geochemical study of soils developed from the unmined Coles Hill uranium deposit (Virginia) was undertaken to determine how phosphorous influences the speciation of uranium in an oxidizing soil/saprolite system typical of the eastern US. This paper presents mineralogical and geochemical results that identify and quantify the processes by which uranium has been sequestered in these soils. It was found that uranium is not leached from the saturated soil zone (saprolites) overlying the deposit due to the formation of a sparingly soluble uranyl phosphate mineral of the meta-autunite group. The concentration of uranium in the saprolites is approximately 1000 mg uranium per kg of saprolite. It was also found that a significant amount of uranium was retained in the unsaturated soil zone overlying uranium-rich saprolites. The uranium concentration in the unsaturated soils is approximately 200 mg uranium per kg of soil (20 times higher than uranium concentrations in similar soils adjacent to the deposit). Mineralogical evidence indicates that uranium in this zone is sequestered by a barium-strontium-calcium aluminum phosphate mineral of the crandallite group (gorceixite). This mineral is intimately inter-grown with iron and manganese oxides that also contain uranium. The amount of uranium associated with both the aluminum phosphates (as much as 1.4 weight percent) has been measured by electron microprobe microanalyses and the geochemical conditions under which these minerals formed has been studied using thermodynamic reaction path modeling. The geochemical data and modeling results suggests the meta-autunite group minerals present in the saprolites overlying the deposit are unstable in the unsaturated zone soils overlying the deposit due to a decrease in soil pH (down to a pH of 4.5) at depths less than 5 meters below the surface. Mineralogical observations suggest that, once exposed to the unsaturated environment, the meta-autunite group minerals react to form U(VI)-bearing aluminum phosphates.

After 30 years of research in the field of nuclear waste disposal, the remaining scientific issues are described. They include long-term monitoring during the reversible period, early coupled processes during the thermal phase, effect of hydrogen generated by corrosion, and volume expansion. The long term behavior of waste packages has been thoroughly investigated, but controversial issues have still to be resolved. In situ investigations concerning the speciation and migration of radionuclides are needed. Concerning the developement of future nuclear systems, a roadmap on waste management issues has to be established.

The disposal of high-level radioactive wastes in a potential geologic repository at Yucca Mountain, Nevada is governed by the U.S. Environmental Protection Agency (EPA) standards and U.S. Nuclear Regulatory Commission (NRC) regulations. The EPA has the responsibility for setting public health and safety and environmental standards for radioactive waste disposal at Yucca Mountain and it is the NRC's responsibility to implement those standards in its regulations to ensure public health and safety and the environment are protected. The U.S. Department of Energy (DOE), as the developer of the potential repository, must submit a license application to the NRC to seek approval to construct the repository. DOE must comply with NRC's regulations for NRC to authorize construction and license operation of a potential repository at Yucca Mountain. In 2005, EPA issued proposed revised standards and NRC issued proposed revised regulations for conducting performance assessment beyond 10,000 years up to 1 million years. The challenge for the EPA and NRC is to develop standards and regulations that provide an appropriate method for evaluating the safety of the potential repository given the unprecedented time period to be analyzed and the inherent uncertainties in estimating the future evolution of the Yucca Mountain site and the containment of the waste in the waste packages. A fundamental aspect of the proposed EPA standard is the specified approach for limiting undue speculation on future behavior of the site by constraining the features, events, and processes that need to be considered in the performance assessment. EPA proposed to limit the assessment of specific features, events, and processes in the period after 10,000 years to effects on the repository system that are most relevant (i.e., ignoring lesser or secondary effects that may add to speculation and uncertainties but would not be expected to have a significant effect on peak dose over a 1 million year period). For example, DOE's performance assessment may (1) limit the analysis of seismic activity to the effects caused by damage to the drifts and the waste package; (2) limit analysis of igneous activity to effects on the waste package that result in release of radionuclides to the atmosphere or ground water; (3) require DOE to include general corrosion in its analysis of engineered barrier performance, and (4) limit the effect of climate variation to those resulting from increased water flowing to the repository. NRC has been reviewing its performance assessment models and techniques to assure they are consistent with EPA's proposed requirements for the period after 10,000 years and adequate to assist the review of a potential license application from the DOE. Currently, only slight modifications to the models and approaches used in the performance assessment for the initial 10,000 years are expected to be needed to accommodate calculations for longer times (e.g., modifications to implement a steady-state value to represent the proposed climate change). NRC staff will be able to use the modified performance assessment to identify additional sensitivities associated with estimating doses over very long time periods and improve its understanding of the performance of a potential repository at Yucca Mountain.

The natural near-field environment in and around the emplacement drifts of the proposed nuclear waste repository at Yucca Mountain, Nevada, includes the host rock, dust, seepage, and pore water. The chemical compositions of these components have been analyzed to provide a basis for assessing possible chemical and mineralogical reactions that may occur after nuclear waste is emplaced. The rock unit hosting the proposed repository has a relatively uniform chemical composition as shown by samples with a mean coefficient of variation (CV) of 9 percent for major elements. In contrast, compositional ranges of underground dust (bulk and water-soluble fractions), pore water, and seepage water are large with mean CVs ranging from 28 to 64 percent for major constituents.

Aqueous dissolution tests of AgI were performed in Na2S solutions in order to evaluate, empirically, dissolution of AgI to release iodine under reducing conditions with sulfide. The results indicated that AgI dissolves to release iodine being controlled by mainly precipitation of Ag2S. However, the dissolution of AgI can be depressed to proceed, and the thermodynamic equilibrium cannot be attained easily. Solid phase analysis for the reacted AgI suggested that a thin layer of solid silver forming at AgI surface may evolve to be protective against transportation of reactant species, which can lead to the depression in the dissolution of AgI.

Chemical extraction techniques and scanning electron microscopy were used to study the distribution and behavior of actinides and rare earth elements (REE) in hydrothermal veins at Adamello, (Italy). The six samples discussed in this paper were from the phlogopite zone, which is one of the major vein zones. The samples were similar in their bulk chemical composition, mineralogy, and leaching behavior of major elements (determined by extraction with 9M HCl). However, there were major differences in the extractability of REE and actinides. The most significant influence on the leaching characteristics appears to be the amounts of U, Th and REE incorporated in resistant host phases. Uranium and Th are very highly enriched in zirconolite grains. Actinides were more readily leached from samples with a higher content of U and Th, relative to the amount of zirconolite. The results show that REE and actinides present in chemically resistant minerals can be retained under aggressive leaching conditions.

Spent nuclear fuel contains fissionable materials (235U, 239Pu, 241Pu, etc.). To prevent nuclear criticality in spent fuel storage, transportation, and during disposal, neutron-absorbing materials (or neutron poisons, such as borated stainless steel, Boral™, Metamic™, Ni-Gd, and others) would have to be applied. The success in demonstrating that the High-Performance Corrosion-Resistant Material (HPCRM) can be thermally applied as coating onto base metal to provide for corrosion resistance for many naval applications raises the interest in applying the HPCRM to USDOE/OCRWM spent fuel management program. The fact that the HPCRM relies on the high content of boron to make the material amorphous – an essential property for corrosion resistance – and that the boron has to be homogenously distributed in the HPCRM qualify the material to be a neutron poison.

The separation of actinides from lanthanides is an important step in the alternative methods for nuclear waste treatment currently under development. Polycyclic molecules containing nitrogen are synthesised and used for solvent extraction. A potential problem in the separation process is the degradation of the molecule due to irradiation or ageing. An addition of nitrobenzene has proved to have an inhibitory effect on degradation when added to a system containing C2-BTP in hexanol before irradiation. In this study, 2,6 di(5,6 diethyl 1,2,4 triazin 3 yl)pyridine (C2-BTP) was dissolved in different mixtures of cinnamaldehyde and hexanol and the effects on extraction after ageing and irradiation were investigated. Similar to nitrobenzene, cinnamaldehyde contains an aromatic ring which generally has a relatively high resistance towards radiolysis. Both C2-BTP in cinnamaldehyde and C2-BTP in hexanol seem to degrade with time. The system with C2-BTP in pure hexanol is relatively stable up to 17 days but then starts slowly to degrade. The solution with pure cinnamaldehyde as diluent started to degrade after only ∼20 hours. The opposite is true for degradation caused by radiolysis; hexanol systems are more sensitive to radiolysis than cinnamaldehyde systems. Most of the radiolytic degradation took place during the first days of irradiation, up to a dose of 4 kGy.

Iron-based amorphous alloys are hard and highly corrosion resistant, which make them desirable for salt water and other applications. These alloys can be produced as powder and can be deposited as coatings on any surface that needs to be protected from the environment. It was of interest to examine the behavior of these amorphous alloys in the standard salt-fog testing ASTM B 117. Three different amorphous coating compositions were deposited on 316L SS coupons and exposed for many cycles of the salt fog test. Other common engineering alloys such as 1018 carbon steel, 316L SS and Hastelloy C-22 were also tested together with the amorphous coatings. Results show that amorphous coatings are resistant to rusting in salt fog. Partial devitrification may be responsible for isolated rust spots in one of the coatings.