Borosilicate glassy materials for immobilization of HLW from Russian WWER (PWR) spent nuclear fuel reprocessing were designed, synthesized in a resistive furnace, and characterized by XRD, SEM/EDS, and FTIR spectroscopy. Chemical durability was determined by PCT-A procedure and compared to EPA glass and reference data. The glasses with 20 and 25 wt.% waste loading were found to be X-ray amorphous, homogeneous and chemically durable. Glass network formally had a relatively low degree of connectedness that was increased due to embedding of different structural groups thus improving chemical durability. Boron is present primarily in trigonal oxygen coordination. The glasses with 40-45 wt.% waste loading contained minor britholite phase concentrating rare earth elements and as expected trivalent actinides. Glassy product with up to 30 wt.% waste loading was also produced by cold crucible inductive melting at the IPCE RAS lab-scale unit equipped with 56 mm inner diameter copper cold crucible and energized from a 10 kW/5.28 MHz generator. The product was composed of vitreous phase and minor britholite with average composition K0.39Sr1.99Fe0.16Nd5.50Si7.96O26.60.

Hydride re-orientation in high burn-up PWR fuel rods was induced by high pressure at high temperatures. The high-burnup specimens were sectioned from PWR rods taken from a 15×15 assembly of the H.B. Robinson (HBR) Unit 2 reactor. Out-of cell benchmark tests performed on unirradiated hydrided Zircaloy-4 specimens were conducted to determine the appropriate temperature, pressure, cooling rate, and number of cooling cycles for the reorientation of the irradiated in-cell specimens. The in-cell hydride reorientation tests were performed using highburnup fuel specimens under a hoop stress ≈145 MPa at 400ºC. The specimens were heated to the target temperature of 400ºC, held for 3 hours, cooled at 1ºC/min to 170ºC,and then heated at 1ºC/min to the target temperature again for five cycles. Post test metallographic examinations showed that a significant amount of radial hydrides were induced in the HBR fuel rods. The length of radial hydride was up to 60 µm. For unirradiated materials, the ductility of the radial hydride treated specimens is significantly reduced as compared to the as-hydrided specimens having the same hydrogen concentration (≈300 wppm in this work). The mechanical testing on irradiated fueled samples with hydride reorientation experiments have been performed, and will be reported separately in the near future.

To classify the chemical form of stable carbon released from unirradiated stainless steel, which is the material used to simulate irradiated stainless steel, under highly alkaline and low-oxygen conditions, type 304 and 316 stainless-steel powders were immersed in 0.005 M NaOH solution. Gas and liquid samples were analyzed to identify the chemical form of carbon released from the stainless steel. The liquid samples were divided into unfiltered and filtered samples. In the gaseous phase, hydrocarbons such as methane and ethane were not detected. In the liquid phase, carboxylic acids (formic and acetic acids) were detected. However, the sum of the carbon concentrations of the carboxylic acids was significantly lower than the total organic carbon (TOC) concentration in the unfiltered samples. In the filtered samples, the TOC concentration was closer to the sum of the carbon concentrations than that for the unfiltered samples. In addition, the concentrations of the metallic elements (particularly Fe and Cr), which are the main constituents of the stainless steels, tended to decrease upon ultrafiltration. This suggests that the sorption of carbon on metallic compounds (e.g., colloidal iron hydroxide) may have occurred.

Types of melter feed materials affect glass production rates. This study focuses on the effects of alumina sources on melting behavior of high-alumina high-level-waste melter feeds containing different alumina sources, namely, gibbsite, boehmite, and corundum. The heat flow from the glass melt to the cold cap, a floating layer of the reacting feed, is partially hindered by a foam layer at the bottom of the cold cap. Volume expansion tests and thermoanalytical methods revealed that a slow-melting feed with corundum foamed extensively, whereas a fast-melting feed with boehmite had a low reaction heat and produced less stable foam. The foam thickness, a critical factor for the rate of melting, estimated using the relationship between the heat conductivity and foam porosity was in reasonable agreement with experimental observation.

A preliminary investigation of the synthesis and characterization of simulant ‘lava-like’ fuel containing materials (LFCM), as low activity analogues of LFCM produced by the melt down of Chernobyl Unit 4. Simulant materials were synthesized by melting batched reagents in a tube furnace at 1500 °C, under reducing atmosphere with controlled cooling to room temperature, to simulate conditions of lava formation. Characterization using XRD and SEM-EDX identified several crystalline phases including ZrO2, UOx and solid solutions with spherical metal particles encapsulated by a glassy matrix. The UOX and ZrO2 phase morphology was very diverse comprising of fused crystals to dendritic crystallites from the crystallization of uranium initially dissolved in the glass phase. This project aims to develop simulant LFCM to assess the durability of Chernobyl lavas and to determine the rate of dissolution, behavior and evolution of these materials under shelter conditions.

The corrosion of unirradiated and unsensitized Advanced Gas-cooled Reactor cladding material, 20/25/Nb stainless steel, was studied as a function of temperature and [Cl-] typical of those found in interim spent fuel storage pond waters. With respect to preventing corrosion, it found to be advantageous to dose the ponds to pH≈11.4. At pH lower than 7, the initiation of pitting is observed at ∼0.4V vs Ag/AgCl, an undesirable condition as pits are considered to be initiators of stress corrosion cracking (SCC) which may contribute to loss of cladding integrity during storage. Such pits are not seen at a pH 11.4 for the expected and projected pond operating temperatures of 24-60°C. There generally appears to be no localised corrosion threat to cladding as the temperature is increased in this range, although substantial pit formation is observed at the extreme maloperation temperature of 90°C at pH 11.4 indicating a loss of protection.

Mechanical damage of non-metallic nuclear wasteforms can be caused by electrical fields induced by decaying clusters of radionuclides surrounded by an insulating matrix. We assess the electric fields near clusters with decaying radionuclides 244Cm, 241Am, 238,239Pu and 137Cs in a glass matrix determining that matrix destruction can gradually occur via electric breakdown discharges and diffusion-controlled change in form of clusters. The most important parameters that control potential matrix destruction are the radioactive cluster (inhomogeneity) size, radionuclide specific radioactivity and effective electrical conductivity of the matrix.

The dynamics of gas evolution during the thermal oxidation of 30% solutions of tri-n-butyl phosphate (TBP) in Isopar-M saturated with 4.3, 8.2 and 12.0 mol/L nitric acid was investigated in open vessels in the temperature range of 70°C to 150°C. The volume of the released gases has been measured. The influence of the pre-irradiation by accelerated electrons has been determined for the thermolysis and accumulation of liquid products of TBP degradation in the system TBP-Isopar-M-HNO3. The temperature borders of oxidation processes transition to autocatalytic regime have been estimated. It has been shown, that the conditions resulting in growth of autocatalytic oxidation do not exist under heating of single-phase systems in open vessels.

This study investigates the use of ‘simple’ glasses, comprising six components, to represent the structure of complex LAW glasses proposed for Immobilized Low Activity Wastes from the Hanford site in the USA. The 18 elements present in ILAW glasses LAW A44, ORP LB2, and LAW A23 were represented by Al2O3, B2O3, CaO, Na2O, SiO2, and ZrO2 according to their coordination chemistry and their roles as network formers and modifiers. The dissolution behavior of each ‘simple’ glass was compared to its corresponding candidate “complex” LAW glass through PCT-B tests. Significant differences were observed; the durability of complex glasses was concluded to be LAW A44 > ORP LB2 ≥ LAW A23 whereas in their simplified versions the order was LAW A44 > LAW A23 > ORP LB2. These results are discussed in relation to compositional differences and highlight the importance of minor glass components in controlling glass durability. The implications of these results for the use of simplified glass compositions are discussed.