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.

Zirconolite glass-ceramics are being developed as potential wasteforms for the disposition of Pu wastes in the UK. Previous studies utilised a variety of surrogates whilst this work uses both cold-press and sinter and hot isostatic press methods to validate the wasteform with PuO2. A cold press and sinter sample was fabricated as part of a validation study for plutonium incorporation in hot isostatically pressed (HIPed) wasteforms. The results confirmed the cold-press and sinter, achieved successful waste incorporation and a microstructure and phase assemblage that was in agreement with those expected of a HIPed equivalent. A HIP sample was fabricated of the same composition and characterised by SEM and XRD. Results were in agreement with the sintered sample and achieved complete waste incorporation into the glass-ceramic wasteform. These samples have demonstrated successful incorporation of PuO2 into glass-ceramic HIPed wasteforms proposed for processing Pu-based waste-streams in the UK.

Cladding tubes of water-cooled nuclear reactors are usually made of Zircaloy and are an important retaining element for radionuclides present in the fuel both during predisposal activities such as reloading of fuel assemblies from interim storage casks to final disposal casks and during final disposal in the case of canister breaching. However, cladding integrity is affected by various processes during reactor operation and beyond, e.g. fuel cladding chemical interaction and fission product precipitation onto the inner cladding surface. Using experimental and modelling methods, the radionuclide inventory of an irradiated Zircaloy-4 plenum section is analyzed. Quantities of 235/238U, 237Np, 238/239/240/241/242Pu, 241/243Am, 243/244Cm besides 14C, 55Fe, 125Sb, 154Eu, and 134/137Cs were (radio-)chemically determined in digested Zircaloy-4 subsamples. Measured inventories of activation products in the Zr-alloy are in good agreement with calculated values. However, amounts of actinides and fission products exceed the calculated inventory by factor ∼57 (minor actinides and non-volatile fission products) and ∼114 (137Cs). Excess of minor actinides and part of enhanced Cs inventory originate from fuel residues deposited on the inner cladding surface during fuel rod fabrication, whereas vast amount of cesium is volatilized from subjacent fuel pellets and transported to the plenum.

In a scenario where ground water enters a canister for spent nuclear fuel in a deep geological repository, the presence of dissolved ions in the water could possibly influence the fuel dissolution due to effects on radiolysis yields. One species of particular interest in this context is bromide, which has a proven ability to scavenge hydroxyl radicals much faster than molecular hydrogen does. As a result, bromide could inhibit the beneficial effect of dissolved hydrogen, which has been shown in γ-radiolysis experiments. However, already a few hundred years after repository closure, α-decay starts to dominate in the radiation field from the spent fuel. Hence, the effects of α-radiolysis are expected to govern the fuel dissolution over the geological timeframes of the repository. In the present work, α-radiolysis experiments have been performed to determine the effect of bromide ions on the yield of hydrogen peroxide by mass spectrometric measurement of its decomposition product oxygen. The use of high activity 238Pu solutions has made it possible to study this effect during pure α-radiolysis from a homogeneously distributed radiation field. To simulate deep bedrock repository conditions, and to minimize the influence of in-leaking O2 from air, the studies were performed using graphite sealed stainless steel autoclaves with an initial atmosphere of 10 bar H2. The results show that addition of 1 mM Br- to the solution gives no significant effect on the O2 yield for radiation doses up to 2 MGy. This lack of effect is most likely explained by the limited radical escape yields from radiation tracks in pure α-radiolysis.

Cerium incorporation in zirconolite glass-ceramic systems, consolidated by hot isostatic pressing, was investigated. Samples were formulated to target Ce incorporation on the Ca and / or Zr sites. Results show successful incorporation of Ce into the ceramic phase although complete waste digestion was not achieved. The formation of a Ce-bearing perovskite phase when targeting Zr substitution, was associated with reduction of Ce4+ to Ce3+. This study is part of preliminary work towards fabricating Pu-bearing glass-ceramic HIP samples.

In disposal of high-level radioactive waste, carbon steel overpack will be corroded after closure of the repository, creating a reducing, low-pH environment around the repository. A plutonium diffusion experiment was performed over 15 years with Kunigel V1, which is a typical Japanese bentonite that contains about 50% montmorillonite, in contact with an iron coupon. A tracer solution (10 µL) containing 1 kBq of 238Pu was applied at the interface between the iron coupon and compacted bentonite that was saturated with deionized water. After the diffusion period, the plutonium distribution in the bentonite specimen was measured with an alpha scintillation counter, and the iron and sodium distributions were obtained by inductively coupled plasma-mass spectrometry. Plutonium penetrated into the bentonite to a depth of 2 mm, and more than 90% of plutonium remained in corrosion product at the interface. The bentonite around the interface was dark green like green rust or magnetite according to visual observation. Iron was detected throughout the bentonite and there was a particularly high iron concentration from the interface to a depth of 2 mm, whereas the sodium concentration decreased slightly from the interface to 2 mm. We proposed that ferrous ions diffused into bentonite as the iron coupon corroded and precipitated such as magnetite. The magnetite precipitation would decrease the bentonite pH, resulting in the dissolution and migration of plutonium. Small cracks were also observed 2 to 6 mm from the interface and could have been formed by the migration of hydrogen generated by corrosion.

The electronic structure of PuB6, an actinide analog of SmB6 , was investigated making use of a combination of the density functional theory (DFT), and the exact diagonalization (ED) of an effective discrete Anderson impurity model. Intermediate valence ground state with the f-shell occupation n4f =5.5 for the Pu atom in PuB6 is calculated. The 5f-shell magnetic moment is completely compensated by the moment carried by the electrons in the conduction band. Already in DFT, PuB6 is an insulator with a small amount of holes near the X-point, and the indirect band gap of ≈60 meV. This band gap becomes direct in DFT+ED calculations supporting the idea of “topological Kondo insulator” in PuB6. Connection between the electronic structure of PuB6 and δ-Pu is established. We propose that these materials belong to a new class of intermediate valence “Racah” materials with the multi-orbital “Kondo-like” singlet ground-state.

Results of characterization of 238Pu-doped Eu- and La-monazites using single crystal XRD, Raman and XAFS spectroscopy and TEM are presented. It is shown that despite significant accumulated doses (up to 9x1018 α-decays/gram) the Eu-monazite remains a single crystal. Unusual foamy structures are observed by TEM and are interpreted as recrystallisation of domains damaged by recoil U-ions. Partial recrystallisation of the surface material is also supported by Raman and luminescence data.

Plutonium metal is a very unusual element, exhibiting six allotropes at ambient pressure, between room temperature and its melting point, a complicated phase diagram, and a complex electronic structure. Many phases of plutonium metal are unstable with changes in temperature, pressure, chemical additions, or time. This strongly affects structure and properties, and becomes of high importance, particularly when considering effects on structural integrity over long periods of time [1]. This paper presents a time-dependent neutron total scattering study of the local and average structure of naturally aging δ-phase 239Pu-Ga alloys, together with preliminary results on neutron tomography characterization.

In disposing of high-level radioactive waste, the drop in pH in the repository as the iron overpack corrodes must be considered. Plutonium migration behavior may be affected by the pH of pore water in compacted bentonite barriers in high-level waste repositories. To examine the effect of pH on migration behavior, H-bentonite was prepared by treating Japanese Na-bentonite, Kunipia-F, with hydrochloric acid. Diffusion experiments were performed with mixtures of Na- and H-bentonites. The pH value in the pore water of the water-saturated bentonite mixtures decreased from 8 to 3 as the mixing ratio of H-bentonite increased. Diffusion experiments were carried out by using 238Pu then apparent diffusion coefficients were determined from the plutonium distribution in the specimens. The apparent diffusion coefficients were on the order of 10-13 to 10-12 m2/s at pH values lower than 4, whereas they were less than 10-14 m2/s at pH values higher than 6.5. These results indicate that plutonium diffused faster as Pu3+ or PuO22+ due to disproportionation at lower pH while plutonium could be retarded as Pu(OH)40 by sorption on bentonite at higher pH.

Samples of 238Pu-doped single-phase ceramics based on cubic zirconia, Zr0.79Gd0.14Pu0.07O1.93, and monazite, La0.9Pu0.1PO4, have been studied by static leach test in distilled water. Before leach test accumulated doses were (in alpha-decays/m3 x 1026): from 1.6 to 1.7 – for cubic zirconia; and 1.0 – for monazite. Despite high radiation damage both phases remained crystalline according to XRD analysis. The results of static leach tests demonstrate the following Pu normalized mass loss (in g/m2, 90°C, 28 days): from 0.3 to 0.7 – for cubic zirconia; and 1.6 – for monazite. These data are discussed in comparison with results of previous leach tests carried out at lower accumulated doses.

To assess the stability of spent fuel in the highly alkaline chemical environment of the Belgian Supercontainer design, static leach experiments were performed with depleted UO2 and 238Pu-doped UO2 at different SA/V ratios for 1.5 years in cement waters (11.7< pH < 13.5) at ambient temperature and under argon atmosphere. The influence of the calcium concentration on the uranium release was also investigated. While the ultrafiltered U(IV) concentration was 10-9-10-8 mol.L-1 and independent of the pH, the U(VI) release from the UO2 surface was enhanced by the OH- concentration, leading to soluble U concentrations up to 10-5 mol.L-1 at high SA/V and under the influence of radiolysis. Together with the high Ca concentration, this can lead to the formation of Ca-U(VI) colloids as precursor of Ca-U(VI) secondary phases, decreasing the soluble U concentration. The precipitation of Ca-U secondary phases was however not clearly evidenced by surface analyses.

Various candidate waste matrices such as nuclear waste glasses, ceramic waste forms and low-specification “storage” MOX have been considered within the current UK geological disposal program for the immobilization of separated civilian plutonium, in the case this material is declared as waste. A review and evaluation of the long-term performance of potential plutonium waste forms in a deep geological repository showed that (i) the current knowledge base on the behavior and durability of plutonium waste forms under post-closure conditions is relatively limited compared to HLW-glasses from reprocessing and spent nuclear fuels, and (ii) the relevant processes and factors that govern plutonium waste form corrosion, radionuclide release and total systems behavior in the repository environment are not yet fully understood in detail on a molecular level. Bounding values for the corrosion rates of potential plutonium waste forms under repository conditions were derived from available experimental data and analogue evidence, taking into account that the current UK disposal program is in a generic stage, i.e. no preferred host rock type or disposal concept has yet been selected. The derived expected corrosion rates for potential plutonium waste forms under conditions relevant for a UK geological disposal facility are in the range of 10-4 to 10-2 g m-2 d-1 and 10-5 to 10-4 g m-2 d-1 for borosilicate glasses, and generic ceramic waste forms, respectively, and ∼5·10-6 g m-2 d-1 for storage MOX. More realistic assessments of the long-term behavior of the waste forms under post-closure conditions would require additional systematic studies regarding the corrosion and leaching behavior under more realistic post-closure conditions, to explore the safety margins of the various potential waste forms and to build confidence in long-term safety assessments for geological disposal.

The density-functional-theory model for plutonium metal is shown to be consistent with recent magnetic measurements that suggest anti-ferromagnetism in Pu-Ga alloys at low temperatures. The theoretical model predicts a stabilization of the face-centered-cubic (fcc, δ) form of plutonium in an anti-ferromagnetic configuration when alloyed with gallium. The ordered magnetic phase occurs because Ga removes the mechanical instability that exists for unalloyed δ-Pu. The cause of the Ga-induced stabilization is a combination of a lowering of the band (kinetic) and electrostatic (Coulomb) energies for the cubic relative to the tetragonal phase.

The migration behavior of plutonium is expected to be affected by the corrosion products of carbon steel in compacted bentonite at high-level waste repositories. Electrochemical experiments were carried out to simulate the reducing environment created by ferrous iron ions in equilibrium with anoxic corrosion products of iron. The concentration profiles of plutonium could be described by the convection -dispersion equation to obtain two migration parameters: apparent migration velocity Va and apparent dispersion coefficient Da. The apparent migration velocity was evaluated within 1 nm/s and was found to be independent of the experiment duration and the dry density of bentonite in the interval 0.8-1.4 Mg/m3. The apparent dispersion coefficient increased with the experiment duration at a dry density of 1.4 Mg/m3. The results for other dry densities also showed the same trend. These findings indicate that plutonium migration likely starts after ferrous ions reach the plutonium, in other words, the reducing environment due to ferrous ions could change the chemical form of plutonium and/or the characteristics of compacted bentonite. The apparent diffusion coefficient was estimated to be around 0.5 to 2.2 µm2/s and increased with decreasing the dry density of bentonite.

A calcium phosphate ceramic waste-form has been developed at AWE for the immobilisation of chloride containing wastes arising from the pyrochemical reprocessing of plutonium. In order to determine the long term durability of the waste-form, aging trials have been carried out at PNNL. Ceramics were prepared using Pu-239 and -238, these were characterised by PXRD at regular intervals and Single Pass Flow Through (SPFT) tests after approximately 5 yrs.

While XRD indicated some loss of crystallinity in the Pu-238 samples after exposure to 2.8 x 1018 α decays, SPFT tests indicated that accelerated aging had not had a detrimental effect on the durability of Pu-238 samples compared to Pu-239 waste-forms.

During this century, humankind must deal with increasing demand for energy and the growing impact of burning fossil fuels. Nuclear power, which presently produces 14% of global electricity, is a low-carbon-emissions alternative. However, the sustainability of nuclear power depends on the amounts of uranium and thorium available, the economics of their recovery from ore deposits, and the safety and security of nuclear materials. Unlike combustion of hydrocarbons, which determines the amount of fuel needed for a given amount of energy, nuclear reactions can create additional fissile isotopes. Hence, the choice of nuclear fuel cycle profoundly affects the size of the nuclear resource, as well as nuclear waste management and the risk of proliferation of nuclear weapons. We argue that uranium resources, identified and yet to be discovered, could sustain increases in nuclear power generation by a factor of two or three through the end of this century, even without advanced closed-fuel-cycle technologies.

Plutonium and Pu-Ga alloys have been observed to have anomalous hydrogen solubility behavior, including a significant concentration dependence of hydrogen diffusivity in the dilute regime, a sharp drop off in the hydrogen solubility constant in the dilute regime, and a near complete absence of change in the Sieverts’ constant as the alloys are heated across phase transformation boundaries. We are investigating the possibility that a vacancy mechanism is responsible for this behavior. X-ray diffraction measurements show a 0.14% lattice contraction in Pu-2 at. % Ga alloys when they are charged with ~2 at. % hydrogen. The lattice re-expands when the hydrogen is removed. Density functional calculations show that increasing the number of hydrogen atoms associated with a vacant lattice site in Pu lowers the energy of the hydrogen-vacancy complex. These observations support the idea that vacancies are stabilized by hydrogen in the Pu lattice well beyond their thermal equilibrium concentration and could be responsible for the anomalous hydrogen response of Pu.

Two LaBS glasses containing 9.5 wt.% (#1) and 5.0 wt.% PuO2 (#2) were prepared by melting in Pt ampoules at 1500 C and examined by scanning electron microscopy with energy dispersive X-ray spectroscopy. The bulk of sample #1, both as-prepared and stored for 3 yrs, was amorphous with homogeneous PuO2 distribution. Sample #2, especially after storage for 2-3 yrs, was partly devitrified primarily in the near-surface area. As followed from X-ray elemental maps, the vitreous phase was enriched with Al and Si whereas larger elongated and smaller dendrite crystals strongly enriched with rare earths (La, Nd, Gd) and Si and minor amounts of Hf may be attributed to britholite. A minor concentration of Pu was also observed in this phase. Moreover, relatively minor amounts of white regular crystals with high PuO2 and lower HfO2 contents were observed in the samples and are probably associated with PuO2 and a PuO2-HfO2 cubic solid solution phase. Nevertheless, even in devitrified areas of the samples, the majority of the Pu remained in the vitreous phase where it was homogeneously distributed.

Simple Hydroxamic acids (XHAs) are salt free, organic compounds with affinities for hard cations such as Fe3+, Np4+, Pu4+ and have been identified as suitable reagents for the control of Pu and Np in advanced nuclear fuel reprocessing. Building upon previous work on the neptunium(IV)-formohydroxamic(FHA) acid system [1], a model that describes the hydrolysis of the acetohydroxamate moiety has been extended to include hydrolysis of bishydroxamatoneptunium(IV) complex. The model has been used to determine the rate constants for hydrolysis of mono- and bis-acetohydroxamatoneptunium(IV) at 25 °C, which were found to be 1.0×10-5 dm3 mol-1 s-1 and 5.0×10-5 dm3 mol-1 s-1, respectively.