We have developed a calcium phosphate ceramic based on the mineral phases chlorapatite, Ca5(PO4)3Cl, and spodiosite, Ca2(PO4)Cl, which, on the basis of evidence from non-active simulant studies, is expected to provide an extremely effective host for immobilizing the chloride constituents resulting from the pyrochemical reprocessing of Pu. The immobilized product is in the form of a free-flowing, non-hygroscopic powder in which the chlorides are chemically combined within the mineral phases. In order to provide a monolithic wasteform suitable for long term storage, the feasibility of encapsulating this product in a compatible glass is being assessed. The final wasteform will be manufactured by a pressureless sintering route with sodium aluminophosphate based glasses currently being developed for this purpose. These glasses are of particular interest due to a combination of useful properties which include good sintering characteristics, together with excellent chemical durability. We have noted, however, that during sintering the Na present in the glass reacts with the chlorapatite and spodiosite phases to form a dispersion of halite crystals within the final wasteform. The formation of free halite within the wasteform is clearly undesirable, but can be minimized by careful control over the processing parameters, in particular the sintering temperature, or by reducing the chloride loading in the wasteform. The implications of these findings are highlighted and discussed.

At present spent nuclear cores from Alfa class submarines are being stored in temporary facilities, designed to sore them only for a few years (3 to 5 years). This paper assesses the problems of nuclear safety during long-term unplanned forced storage of spent fuel because of the delay in building permanent facilities. The Nuclear Submarines (NS) cores use heavy liquid metal lead-bismuth coolant in the primary circuit of nuclear reactor. The radiation parameters of the discharged cores from NS of the Project 705K (this is the number of ship engineering design) are determined. Based on that determination a step-by-step solution is proposed and the technological and other challenges that the Project faces are considered.

Argonne National Laboratory has developed an electrometallurgical treatment for DOE spent metallic nuclear fuel. Fission products are immobilized in a durable glass bonded sodalite ceramic waste form (CWF) suitable for long term storage in a geological repository. Cesium is estimated to be in the waste form at approximately 0.1 wt.%. The exact disposition of cesium was uncertain and it was believed to be uniformly distributed throughout the waste form. A correlation of X-ray diffractometry (XRD), electron microscopy (EM), and nuclear magnetic resonance spectroscopy (NMR) performed on surrogate ceramic waste forms with high cesium loadings found a high cesium content in the glass phase and in several non-sodalite aluminosilicate phases. Cesium was not detected in the sodalite phase.

A Titanium Grade 7 (Ti-7) drip shield is being considered for use in the potential nuclear waste repository at Yucca Mountain, Nevada. The purpose of the drip shield is to keep water from reaching the surfaces of the canisters containing nuclear waste. Among the species that may be expected to be present in the water are heavy metals. This paper explores the effects of five species, lead, mercury, cadmium, tin and sodium, on the dissolution rate of Ti-7. The results of these tests indicated that lead, mercury, cadmium and tin ions do not significantly accelerate the general corrosion of Ti-7 under the conditions studied. A highly concentrated solution of NaOH was found to cause significant dissolution of Ti-7. Such a solution is a reasonable approximation of possible service environments.

The enthalpy of melting and the heat capacity of liquid zirconolite (CaZrTi2O7) are estimated as equal to 200 ± 20 kJ·dmol−1 and 350 ± 50 J·mol−1·K−1, respectively. Thermodynamic functions of solid and liquid zirconolite are calculated based on these estimated data and the results of Navrotsky et al. On the basis of these thermodynamic functions, computational thermodynamic simulation is performed on the thermochemical synthesis of zirconolite-bearing materials. Demonstration indicates that synthesis of zirconolite-like matrix materials is possible using the self-sustaining exothermic reaction.

An oscillatory-zoned single crystal of zircon (ZrSiO4) from Sri Lanka (560 Ma) provides a unique opportunity to study radiation effects over the dose range of 2.1 - 10.6·1015 á-decay events/mg (0.1 - 0.5 dpa). Each zone has a different radiation dose due to variations in the U (0.09 - 0.53 wt %) and Th (0 - 0.11 wt %) content, but all zones share the same thermal history. The zones span a nearly complete range of microstructures from crystalline to aperiodic. Measurements of birefringence show an inverse-linear correlation with dose over the range of doses observed. Micro-Raman spectroscopy results show that a decrease in peak frequency relative to an undamaged zircon, peak broadening and decreases in peak intensity occur with the accumulation of radiation damage.

IE-911 is a bound form of cystalline silicotitanate (CST) that was extensively tested for removing 137Cs from the Savannah River Site (SRS) tank wastes. In some simulant tests, column plugging incidents were observed, which led to thorough investigations to determine the causes and to develop protocols to avoid future plugging incidents. A related problem was the apparent decrease in Cs scavenging capability in some long-term tests. Our studies revealed that the interaction of IE-911 with the highly basic, high ionic strength, SRS average salt simulant could result in precipitation of; 1) poorly crystalline Nb-oxide, or 2) aluminosilicate zeolitic phases. The source for the Nb-oxide precipitate was determined to be a minor impurity phase that is a byproduct of CST manufacturing. The mechanisms of dissolution and re-precipitation of this phase in column pretreatment solution were investigated, and a protocol to rid IE-911 of this impurity was devised. The source material for the aluminosilicate zeolite precipitate was determined to be predominantly from the waste solution rather than the IE-911. Solubility experiments coupled with a thermodynamic analysis provided a protocol to predict when aluminosilicate precipitation will and will not occur. Finally, it was also established that aluminosilicate precipitation on the surfaces of the IE-911 granules could also account for an apparent decrease in equilibrium Kd and decrease in kinetics of Cs sorption.

Carbon steel is one of the candidate overpack materials for high-level waste disposal and is expected to assure complete containment of vitrified waste glass during an initial period of 1000 years in Japan. The lifetime of the carbon steel overpack will depend on its corrosion rate. The corrosion rate of carbon steel is reduced by the presence of buffer material such as bentonite. Buffer material will delay the supply of corrosive materials and discharge of corrosion products through it. Carbon steel overpack will be corroded by consuming oxygen introduced by repository construction after closure of repository and then will keep the reducing environment in the vicinity of repository. The reducing condition will be expected to retard the migration of redox-sensitive radionuclides by lowering their solubilities. Therefore, the diffusion of corrosion products of iron in buffer material is important to discuss the corrosion rate of overpack, migration of redox-sensitive radionuclides and properties of buffer material. The purpose of this paper is to study diffusion behavior of a corrosion product of iron in compacted bentonites under a reducing condition with a carbon steel. The diffusion mechanism of iron in the compacted bentonites were discussed by estimation of iron species in the bentonite pore water. There were two diffusion paths of iron in the compacted bentonites used in this study; the fast path has low capacity of iron, ca. 1wt%, and large apparent diffusion coefficient, ca. 10−12 m2/s and the slow path has high capacity of iron, ca. 10wt%, and small apparent diffusion coefficient, ca. 10−14 m2/s.

Effects of alpha decay on the ceramic waste form used to immobilize fission products and actinides accumulated from electrometallurgical treatment of Department of Energy metallic spent nuclear fuel is being studied. This involved the preparation of a surrogate ceramic waste form containing two-weight percent plutonium-238 and monitoring changes to the waste form resulting from alpha-decay damage. The phase assembly of the ceramic waste form consists of glass-bonded sodalite with small quantities of nepheline, halite and plutonium dioxide. The waste form has been monitored for three years and has acquired an internal alpha-decay dose of 1 × 1018 disintegrations per gram of material. Methods used to monitor the material include: density determination, chemical durability measured by immersion testing, microscopy and powder xray diffraction. The investigation has, to date, found little alteration to the waste form due to alpha-decay damage. X-ray diffraction analysis has detected a unit cell volume increase of 0.7 percent of the plutonium dioxide phase. Furthermore, bubbles and/or voids have been observed by transmission electron microscopy in the sodalite and glass phase. The origin of these defects is unclear and under investigation.

Discrepancies between classical kinetic rate law theory and experiment were quantitatively assessed and found to correlate with macromolecular amorphous separation in the sodium borosilicate glass system. A quantitative reinterpretation of static corrosion data and new SPFT data shows that a recently advanced protective surface layer theory fails to describe the observed dissolution behavior of simple and complex silicate glasses under carefully controlled experimental conditions. The hypothesis is shown to be self-inconsistent in contrast with a phase separation model that is in quantitative agreement with experiments.

Depleted uranium dioxide (DUO2) waste packages (WPs) for disposal of spent nuclear fuel (SNF) are being investigated to (1) reduce radionuclide releases from WPs, (2) decrease the potential for repository nuclear criticality events, (3) provide radiation shielding, and (4) provide a means to beneficially use excess depleted uranium (DU). The DUO2 is incorporated into the WP as (1) a particulate fill for void spaces within the package and (2) a component of a DUO2-steel cermet (DUO2 embedded in steel) that replaces the steel components of the WP. Depending upon the design, there is 3 to 8 times as much DUO2 as SNF UO2 in the WP. Most radionuclides in the SNF cannot be released until the UO2 crystal structure is destroyed. The DUO2 surrounding the SNF slows the degradation of the SNF UO2 in the interior. This behavior is similar to the mechanisms that slow the degradation of natural uranium ore bodies containing UO2. The results of initial investigations and the expected thermodynamic WP behavior are described.

Inorganic ion exchange media typically exist as fine powders, making large-scale use impractical, unless the media can be affixed to an appropriate matrix. Likewise, organic chelating agents are typically dissolved in a solvent and absorbed into porous matrices for use in extraction chromatography. The most common matrices utilized in both cases are organic materials, that are not compatible with high radiation fields or acceptable as final waste forms. Recent investigations have shown that ion exchange sorbents can be effectively loaded within a porous crystalline silica (Gubka) matrix. This approach allows for target radionuclides to be adsorbed into a porous micro-crystalline glass matrix which encapsulates the contaminant and becomes the final waste form. Subsequent to adsorption of the radionuclides, the Gubka matrix can be compressed in a hot uniaxial press, resulting in an even greater volume reduction. The porous glass matrix is produced in Russia using fly ash residue from coal combustion power generating plants. It consists of consolidated arrays of hollow glass cenospheres and is termed Gubka which is the Russian word for sponge. This paper describes results of a collaborative research program between the Khlopin Radium Institute, St.Petersburg, Russia, the Institute of Chemistry and Chemical Technologies, Krasnoyarsk, Russia, the Mining and Chemical Combine, Zheleznogorsk, Russia, and the Idaho National Engineering and Environmental Laboratory. Ammonium molybdophosphate (AMP) for the removal of cesium from acidic liquid waste has been successfully incorporated into Gubka matrices. Test results for cesium removal, using AMP-Gubka, are discussed.

Yucca Mountain, Nevada, is under consideration as a potential site for a repository for high-level radioactive waste. Total-system performance-assessment simulations are performed to evaluate the safety of the site. Features, events, and processes have been systematically evaluated to determine which ones are significant to the safety assessment. Computer models of the disposal system have been developed within a probabilistic framework, including both engineered and natural components. Selected results are presented for three different total-system simulations, and the behavior of the disposal system is discussed. The results show that risk is dominated by igneous activity at early times, because the robust waste-package design prevents significant nominal (non-disruptive) releases for tens of thousands of years or longer. The uncertainty in the nominal performance is dominated by uncertainties related to waste-package corrosion at early times and by uncertainties in the natural system, most significantly infiltration, at late times.

XANES and EXAFS data were collected and analyzed to characterize vanadium in borosilicate glasses used for immobilization of sulfur-containing nuclear wastes. Earlier studies suggested that adding vanadium to the melt improves sulfur solubility. Data are presented for a variety of borosilicate glasses, some containing sulfur and some sulfur-free, that have V2O5 concentrations as high as 12 wt%, and for crystalline vanadium sulfide, silicate, and oxide standards. The data for all glasses investigated indicate that most or all vanadium has a +5 valence and is tetrahedrally coordinated by oxygen atoms. Both XANES and EXAFS also show that glasses synthesized under reducing conditions can have pentacoordinated V+4 populations up to approximately 20 to 25% of all vanadium present with the remainder being V+5O4. There is no evidence from XANES or EXAFS of V-S bonds in any of the glasses investigated.

The conductivity of hygroscopic salt deposits containing Na+, K+, NO3– and Cl– ions was measured in air as a function of relative humidity at constant temperatures. The deliquescence relative humidity (DRH) of multicomponent salts containing Na+, K+, NO3– and Cl– was also determined experimentally. The results of the conductivity experiments show that the conductivity of initially dry salt deposits start to increase after reaching a relative humidity value that is 15 to 20% lower than the DRH of the salt. When the DRH is reached, the conductivity increases dramatically as the salt dissolves and transforms into a saturated aqueous phase. The increase in conductivity at humidities below the DRH is attributed to the adsorption of water on the surface of the salt particles. Because of the increase in conductivity, the initiation of aqueous corrosion of metals in contact with hygroscopic salts may occur at a relative humidity much lower than the DRH of the salt. Thus, the onset of aqueous corrosion of metallic nuclear waste package and the drip shield may be earlier, the duration may be longer, and the temperature at which it occurs may be higher than assumed based on the DRH of the salt. The results of the DRH experiments show that the DRH of a salt mixture is usually significantly lower than that of any of its component pure salt.

Two high burnup BWR fuels [64 and 71 (MWd)/kgU], one of which contained 2% Gd, and two low burnup PWR fuels [30 and 45 (MWd)/kgU], are tested by dripping groundwater on the fuel under oxidizing and hydrologically unsaturated conditions for 2.4 and 8.2 yr, respectively, at 90°C. The 99Tc, 129I, 137Cs, 97Mo, and 90Sr releases are presented to show the effects of reaction time and of gadolinium (Gd) on nuclide release. For the PWR fuels, the five nuclides have similar fractional release rates at 8.2 yr. For the BWR fuels, the presence of 2% Gd reduced the 99Tc cumulative release fraction after 2.4 yr by about an order of magnitude from that of a fuel with a similar burnup but without added Gd.

We performed a series of dissolution experiments with well-characterized pyrochlore ceramics with the formula A2Ti2O7, where A = Y3+, Gd3+, or Lu3+ in H2O- and D2Obased solutions [pH(D) = 2] at 90°C. Normalized log10 dissolution rates (g·m−2·d−1) in H2O-based solutions increase from Lu2Ti2O7 (−3.2 to –3.3) to Gd2Ti2O7 (−2.6 to –2.9), to Y2Ti2O7 (−1.9 to –2.0). Rates in D20-based solutions are indistinguishable from rates in H2O, indicating that release of elements is probably not diffusion controlled. A recent dissolution model, based on ligand-exchange reactions, suggests that the rate of reaction should increase in inverse order of the cation field strength: Lu < Y < Gd (where the cation denotes the appropriate pyrochlore composition), which is not observed. Evaluation of the thermodynamic stability of the three solids was performed using a linear free-energy model and reported free energies of formation. The calculations indicate that reactivity should follow in the progression Lu < Gd < Y, as observed in the dissolution experiments. Our data indicates, therefore, that dissolution models based on ligand-exchange reactions may not be strictly applicable to simple pyrochlore minerals.

The Tono uranium deposit (Gifu Prefecture, central Japan) was formed about 10 million years ago and available evidence suggests that this deposit has remained stable, i.e. uranium has not migrated significantly since then. Substantial field and modeling work is being carried out to better understand the hydrogeological and geochemical processes affecting or governing uranium transport and immobilization in and around the Tono uranium deposit. A systems analysis approach has been applied to the Tono deposit natural analogue project. Application of a systems analysis approach to a generic research project is novel and is different from prevailing applications. Previous applications of a systems analysis approach have been principally concerned with performance and safety assessments for radioactive waste disposal. The systems analysis approach provides a convenient framework for this project, having been used in the Tono Natural Analogue Project (TAP) to:

• define the system to be studied;

• identify features, events and processes (FEPs) associated with this system;

• identify external factors (EFEPs) which are likely to have influenced the Tono area or study site or uranium deposit;

• define interactions between pairs of FEPs; and

• develop credible scenarios for evolution of the area or uranium deposit.

The results from the systems analysis approach allow us to:

• develop possible scenarios for how the Tono area or study site evolved,

• evaluate uncertainty and how it propagates backwards in time,

• link field characterization data with predictive modeling, and

• identify gaps in data and, hence, identify additional characterization needs.

Crystalline ceramic materials based on the cubic zirconia structure have been proposed as candidate waste forms for the immobilization of weapons grade Pu and other actinides. To evaluate a resistance of these materials to self-irradiation for extended period of time, polycrystalline samples of gadolinia-stabilized cubic zirconia, (Zr,Gd,Pu)O2, doped with approximately 10 wt.% 238Pu were synthesized and characterized. Ceramic synthesis was done by sintering in air at 1500°C for 4 hours using starting precursor materials based on coprecipitated and then calcined oxalates of Zr, Gd, Pu. No differences were observed among the zirconia X-ray diffraction patterns that were obtained immediately after ceramic synthesis, or at 88 and 201 days later. MCC-1 leach tests were performed on ceramic specimens in deionized water at 90°C for 28 days. It was found that without correction for ceramic porosity the initial Pu mass loss (NL) was 0.04 g/m2. This increased to 0.35 and 0.37 g/m2, respectively, 180 and 260 days later. Results obtained allow us to confirm that actinide-doped cubic zirconia is highly resistant to accelerated self-irradiation and therefore, is an efficient material for actinide immobilization in deep geological repositories.

Radiation durability of garnet [A3B2(XO4)3; Ia3d; Z=8] has been examined by 1.0 MeV Kr2+ irradiation with in situ transmission electron microscopy over the temperature range of 50 to 1070 K. The targets were five synthetic garnets incorporating various contents of actinides and andradite, Ca3Fe2Si3O12. The synthetic garnets were silicates (N series) and ferrate-aluminate series (G series).

The critical amorphization temperatures (Tc), above which amorphization does not occur, were determined to be 1050 K for N77, 1130 K for N56, 1100 K for G3, 890 K for G4 and 1030 K for andradite. Tc of the synthetic garnets increased as the average atomic mass of the garnet increased. The maximum transferred energy by ballistic interaction was positively correlated to the atomic mass. The larger cascade size that formed due to the larger Emax might lead to the higher Tc.