Zirconium alloys are common fuel claddings in nuclear fission reactors and are susceptible to the effects of hydrogen embrittlement. There is a need to be able to detect and image hydrogen at the atomic scale to gain the experimental evidence necessary to fully understand hydrogen embrittlement. Through the use of deuterium tracers, atom probe tomography (APT) is able to detect and spatially locate hydrogen at the atomic scale. Previous works have highlighted issues with quantifying deuterium concentrations using APT due to complex peak overlaps in the mass-to-charge-state ratio spectrum between molecular hydrogen and deuterium (H2 and D). In this work, we use new methods to analyze historic and simulated atom probe data, by applying currently available data analysis tools, to optimize solving peak overlaps to improve the quantification of deuterium. This method has been applied to literature data to quantify the deuterium concentrations in a concentration line profile across an α-Zr/deuteride interface.

Grain refinement has been applied to enhance the materials strength for miniaturization and lightweight design of nuclear equipment. It is critically important to investigate the low-cycle fatigue (LCF) properties of grain refined 316LN austenitic stainless steels for structural design and safety assessment. In the present work, a series of fine-grained (FG) 316LN steels were produced by thermo-mechanical processes. The LCF properties were studied under a fully reversed strain-controlled mode at room temperature. Results show that FG 316LN steels demonstrate good balance of high strength and high ductility. However, a slight loss of ductility in FG 316LN steel induces a significant deterioration of LCF life. The rapid energy dissipation in FG 316LN steels leads to the reduction of their LCF life. Dislocations develop rapidly in the first stage of cycles, which induces the initial cyclic hardening. The dislocations rearrange to form dislocations cell structure resulting in cyclic softening in the subsequent cyclic deformation. Strain-induced martensite transformation appears in FG 316LN stainless steels at high strain amplitude (Δε/2 = 0.8%), which leads to the secondary cyclic hardening. Moreover, a modified LCF life prediction model for grain refined metals predicts the LCF life of FG 316LN steels well.

FeCrAl alloys are among the best and most mature accident tolerant fuel cladding candidates produced to date, due to their superior combination of mechanical properties and stability at elevated temperatures. For fuel cladding applications, these materials are drawn into tubes with plugs welded to the ends. The mechanical properties of such welds and the impact on cladding performance have not been fully investigated. A novel mesoscale mechanical test and a variety of microscale tests were performed to evaluate a range of properties including nanoindentation hardness, compression and shear yield strengths, and tensile strengths and elongations. Micromechanical testing generally matched the trends of the larger mesoscale testing, with nanoindentation reproducing the trend the best, although some discrepancies existed in regions with low dislocation content. Mesoscale tensile testing showed good correlation with macroscale tests and revealed that the plug heat-affected zone possessed the lowest strength and ductility. This indicated that failure would occur first in or near this region.

Mass attenuation coefficients (MACs) of Th, U, Np, and Pu for oxygen X-rays have been experimentally determined using an electron microprobe. The MACs were obtained by measuring relative X-ray intensities emitted from ThO2, UO2, NpO2, and PuO2 targets, for incident electron energies from 5 to 30 keV, and processing them with the help of the computer program XMAC. The accuracy of the measured MACs is estimated to be better than 5%. Results are compared with MAC tabulations commonly used in electron probe microanalysis as well as with theoretical photoionization calculations. It is concluded that the MACs implemented in the Monte Carlo simulation program PENELOPE which are based on the photoionization cross-section calculations of Sabbatucci & Salvat [(2016). Theory and calculation of the atomic photoeffect. Rad Phys Chem121, 122–140], provide the best agreement with our measurements. The use of different MAC schemes for the analysis of mixed actinide oxide materials is discussed.

Uranium–35 wt.% zirconium (U–35 wt.% Zr) alloy was annealed for 1 h and 24 h at 650 °C and characterized to understand the early-stage microstructure evolution. Dendritic microstructure with fine (∼300 nm in length) α-U precipitates clustered between dendrite branches were observed in the 1-h annealed sample. After 24-h annealing at 650 °C, the α-U precipitates coarsened, and the dendritic microstructure disappeared because of microstructure homogenization. Furthermore, microchemical homogenization observed with energy-dispersive X-ray spectroscopy analysis suggests that α-U precipitates are approaching thermodynamic equilibrium in the 24-h annealed sample. The findings from this study have potential impacts on the manufacturing and computer modeling of metallic nuclear fuel.

The global expansion of nuclear energy will generate increasing quantities of waste with low levels of plutonium or other nuclear materials (NM) potentially subject to International Atomic Energy Agency (IAEA) safeguards. Reducing requirements on retained wastes has the potential to reduce future demands on already strained IAEA resources. We describe an effort to help the IAEA and Member States better estimate projected waste loads and associated safeguards obligations by developing a reporting tool to estimate types and sizes of future waste-storage and -disposal facilities. States can use such information to plan waste facilities, including size and type. The IAEA can use these data for inclusion in multiple agency reports and products for the benefit of Member States.

Computational modeling is an important aspect of the research on nuclear waste materials. In particular, atomistic simulations, when used complementary to experimental efforts, contribute to the scientific basis of safety case for nuclear waste repositories. Here we discuss the state-of-the-art and perspectives of atomistic modeling for nuclear waste management on a few cases of successful synergy of atomistic simulations and experiments. In particular, we discuss here: (1) the potential of atomistic simulations to investigate the uranium oxidation state in mixed-valence uranium oxides and (2) the ability of cementitious barrier materials to retain radionuclides such as 226Ra and 90Sr, and of studtite/metastudtite secondary peroxide phases to incorporate actinides such as Np and Am. The new contribution we make here is the computation of the incorporation of Sr by C-S-H (calcium silicate hydrate) phases.

Leaching results to compare the dissolution behavior of a new type of fuel with additives (Advanced Doped Pellet Technology, ADOPT) with standard UO2 fuel are presented. Both fuels were irradiated in the same assembly of a commercial boiling water reactor to a local burnup of ∼58 MWd/kgU. Fuel fragments are leached in simplified groundwater in two autoclaves under hydrogen atmosphere, representing conditions in a canister failure scenario resulting in water intrusion for a spent nuclear fuel repository. Preliminary results indicate the uranium concentration decreased to 3-4x10-8 M after 421 days, slightly above the solubility of amorphous UO2. Xe has been detected in the gas phase of both autoclaves. The concentration of Cs and I seems to gradually approach constant values, yet the redox sensitive elements continue to slowly increase with time. The preliminary data obtained supports the hypothesis that there is no major difference in leaching behavior between the two fuels.

This study seeks to understand the low temperature reactions of the salt phase that occur during the vitrification of Hanford Low Activity Waste (LAW). Salts (such as nitrates, sulfates, carbonates, halides, etc.) play a key role in these low temperature reactions as they sequentially melt, decompose, and volatilize during batch-to-glass conversion. To further understand these complex processes, simplified LAW melts containing oxyanion salts (sodium salts of carbonate, sulfate, and/or nitrate) and early melting glass formers (boric acid) have been evaluated using thermal analysis, infrared absorption spectroscopy, and X-ray diffraction. Results from this study indicate that the volatilization behavior of particular salts is influenced by the presence or absence of other salts. NaNO3 volatilization is decreased by the presence of Na2SO4. The addition of either Na2SO4 or NaNO3 to the system may enhance the volatilization of Na2CO3. In all cases, Na2SO4 was retained after melting and was often found to be in two different crystalline phases upon quenching.

In a deep geological repository (DGR) scenario, uranium oxidized in aqueous systems will be stabilized as UO22+ (hexavalent uranium), as a consequence of tetravalent uranium oxidation by radiolytic byproducts. Uranyl cationic species (UO22+) in different speciation forms are expected to be found at the whole pH range conditions. The importance of UO22+ lies in its potential incorporation of trace radioelements onto secondary uranyl phases. In view of the difficulty of U chemistry in natural groundwater, it is necessary to improve speciation assessment techniques so as to understand chemical processes. Raman spectroscopy has been shown as a powerful tool to analyze the speciation of various actinyl (UO22+,NpO2+ and PuO22+) and to determine the distribution of those elements which are more likely to be stable in a near-field groundwater environment. Therefore, the aim of this work is to follow UO22+ changes as a consequence of γ radiation in aqueous media under DGR conditions, and to understand the behavior of UO22+ as a function of aqueous media, which help to understand and predict the potential precipitation of the solid phases formed. In this work, the use of Raman spectroscopy adapted to the empirical analysis of different nuclear applications for initial uranium concentrations 0.04M at ambient atmosphere is shown, i.e. as monitoring tool for UO22+ precipitation as a function of pH, studying UO2(NO3)2·6H2O stability in aqueous solutions representative of groundwater, in particular at ionic strength I = 0.02 – 0.4 M and pH from 7 to 13.2; and to evaluate the effect of γ radiation fields. At 10−4-10-3 M of radiolytically formed H2O2 concentration, the amount of uranium in solution decreased, as a results of secondary phases precipitation. The results obtained will be useful to the performance assessment studies of the Spent Nuclear Fuel (SNF) stored in DGRs. The work performed provides a partial picture of secondary phase formations, as a result of corrosion of SNF in a DGR.

The dissolution of the United Kingdom’s vitrified high-level-waste simulant, CaZn MW28, was investigated following the Product Consistency Test-B protocol for 112 d at 90 °C and in ultra-high-quality water. Residual rate dissolution (stage II) and rate resumption (stage III), after 28 d, was observed. Thermodynamic modelling suggested that solutions were saturated with respect to Mg- and Zn-bearing phases, and the presence of Mg- and Zn-smectite clays was tentatively observed. The formation of these phases was concurrent with a significant increase in the dissolution rate, similar to Stage III behavior seen in other nuclear waste simulant glass materials, indicating that the addition of Mg and Zn to high-level-waste glass (7.3 wt. % combined) significantly influences the dissolution rate.

Improved budgeting of heat loads arising from radiogenic heating in high level wastes (HLW) could allow enhanced usage of geological disposal facility space. Separation of high heat generating nuclides from HLW, such as Cs, would simplify management of heat loads. A potential host matrix for Cs-disposal is hollandite. The incorporation of Cs into the hollandite phase is aided by substitution of cations on the B-site of the structure; these ions may include Ni and Zn. Two series of hollandites, Ni-substituted and Zn-substituted, were synthesised via an alkoxide-nitrate route and consolidated by cold uniaxial pressing and sintering or by hot isostatic pressing. Characterisation of the resultant material by X-ray diffraction and scanning electron microscopy found that hollandite was formed for all levels of substitution. Materials produced via HIP were found to be denser indicating lower Cs loss. HIPed Ni hollandites were found to contain fewer secondary phases and it was concluded that they were the most suitable candidates

Salt cake radioactive waste is a remnant solid salt concentrate after deep evaporation of radioactive evaporator concentrate at WWER NPP’s. The traditional cementing of borate-containing liquid radioactive waste, to which the salt cake belongs, leads to a significant increase in the volume of the final product. This work describes borosilicate vitreous wasteforms developed to immobilize radioactive salt cake waste and comprises data on both glass synthesis and characterization. The composition of glass selected for the purpose of immobilisation of the salt cake radioactive waste allows to include up to 40 wt. % of the oxides contained in the salt cake and to reduce the volume of the final product by more than 2 times compared with the cement compound. The batches were melted in a cold crucible melter at 1200 °C. The normalized cesium leaching rate of the vitrified wasteform product was within range 3.0·10-5 – 3.7·10-6 g/(cm2·day).

Recent efforts have been made toward the integration of the back-end of the nuclear fuel cycle in the United States. The back-end integration seeks to address several management challenges: 1) current storage practices are not optimized for transport and disposal; 2) the impact of interim storage on the disposal strategy needs to be evaluated; and 3) the back-end is affected by—and affects—nuclear fuel cycle and energy policy choices. The back-end integration accounts for the various processes of nuclear waste management—onsite storage, consolidated storage, transport and geological disposal. Ideally, these processes should be fully coupled so that benefits and impacts can be assessed at the level of the full fuel cycle. The paper summarizes the causes and consequences of the absence of integration at the back-end of the nuclear fuel cycle in the U.S., critically reviews ongoing integration efforts, and suggests a framework that would support the back-end integration.

X-ray Absorption Spectroscopy is a technique of fundamental importance in nuclear waste management, as an element specific probe of speciation, which governs radionuclide solubility, immobilisation and migration. Here, we exploit recent developments in laboratory instrumentation for X-ray Absorption Spectroscopy, based on a Rowland circle geometry with a spherically bent crystal analyser, to demonstrate speciation in prototype ceramic and glass-ceramic waste forms. Laboratory and synchrotron XANES data acquired from the same materials, at the Ce and U L3 edges, were found to be in excellent quantitative agreement. We establish that analysable laboratory XANES data may be acquired, and interpreted for speciation, even from quite dilute absorber concentrations of a few mol%, albeit with data acquisition times of several hours. For materials with suitable absorber concentrations, this approach will enable routine element specific speciation studies to support rapid optimisation of radioactive waste forms and analysis of radiological materials in a purpose designed laboratory, without the risk associated with transport and manipulation at a synchrotron radiation facility.

The electric industry is transitioning to higher penetrations of renewables. Hundred per cent renewable penetration is no longer a pipe dream. Rather than by doubling down on existing renewable technologies, we can achieve it by cohesively focusing on the ‘needs’ and working on regulation (regulation should focus on holistic grid needs), operations (e.g., markets and balancing authority products), and innovation (e.g., newer technologies like hydrogen).

The addition of Cr2O3 to modern UO2 fuel modifies the microstructure so that, through the generation of larger grains during fission, a higher proportion of fission gases can be accommodated. This reduces the pellet-cladding mechanical interaction of the fuel rods, allowing the fuels to be “burned” for longer than traditional UO2 fuel, thus maximising the energy obtained. We here describe the preparation of UO2 and Cr-doped UO2 using Hot Isostatic Pressing (HIP), as a potential method for fuel fabrication, and for development of analogue materials for spent nuclear fuel research. Characterization of the synthesised materials confirmed that high density UO2 was successfully formed, and that Cr was present as particles at grain boundaries and also within the UO2 matrix, possibly in a reduced form due to the processing conditions. In contrast to studies of Cr-doped UO2 synthesised by other methods, no significant changes to the grain size were observed in the presence of Cr.

Millimetre UO2 single crystals were cut and oriented at JRC Karlsruhe. The orientation of each face of the parallelepiped single crystals was determined with Laue diffraction and the corresponding surface area by geometric measurements. Then, the (111), (100), (110) faces of each single crystal were polished to optical grade and characterized by XRD in order to confirm the surface orientation. The dissolution of the three single crystals was achieved in nitric acid media under dynamic conditions, at room temperature. Two dissolution regimes were observed for all samples. The normalized dissolution rate measured in the first step was not influenced by the crystallographic orientation of the faces. However, during the second step, (110) oriented faces were found to dissolve 4 times faster than the (100) faces. One explanation could involve the atomic composition of each oriented surface in the fluorite-type structure

The oxide films formed on spent fuel claddings are regarded as a potential source of the instantaneous release of radionuclides, such as 14C, after waste disposal. We investigated the instant release fraction using the irradiated oxide exfoliated from a Zircaloy-2 water rod, whose bundle burnup was 53.0 GWd/MTU. We performed a rapid leaching test in a dilute NaOH solution (pH of 12.5) for 10 min in an ultrasonic bath to ensure the release of radionuclides. The activity ratios of the leached amount to the total amount for 14C, 60Co, and 125Sb were extremely low at approximately 10-4 to 10-3, among which the maximum value was 2.65 × 10-3 for 125Sb. These ratios were higher than that predicted from the thermodynamic solubility of ZrO2, i.e., less than 10-6. However, given the low ratios, it is too conservative to regard the inventory of all radionuclides in the Zircaloy oxide as instantaneous release. A small part of the released 14C was found as volatile species.

Plasma incineration might be a promising technique for the conditioning of various radioactive waste streams. Assessing the long-term durability of the plasma slag is essential to predict its performance during long-term disposal. In this paper, the stability of six plasma treated surrogate cemented concentrates or resins in a high pH environment is investigated. The slags were crushed (2 different granulometries) and immobilized in a cement matrix, after which samples were submitted to long-term durability tests (stability under water at 20 °C; stability in a high relative humidity environment at 38 °C) and to an accelerated Alkali-Silica-Reaction (ASR) test (1 M NaOH at 80 °C). The first results show that the expansion and strength loss of the cement-slag mixtures remain limited in the test conditions, although differences between the different materials and granulometries could be perceived. No visual damage was observed. Some tests are still ongoing and will last 2 years.