High Waste Loading Research Articles

Effect of 241Am buildup during spent fuel cooling on the decay heat of vitrified high-level waste and post-closure safety assessment

Long-term interim storage of spent fuel is a key factor affecting the thermal properties of vitrified high-level waste (HLW). In this study, we observed that increasing the spent fuel cooling-off period from 4 to 100 years decreased the decay heat of vitrified HLW and allowed a volume reduction of up to 13% based on the loading of the waste up to the thermal constraint of the repository. Contrastingly, the glass surface temperature 1000 years after disposal and the temperature-dependent glass dissolution rate increased with the cooling-off period due to 241Am buildup from 241Pu within that period. However, these dissolution rates can be regarded as nearly equivalent throughout the lifetime of glass dissolution because the glass temperature approached the host rock temperature and remained at this temperature for most of its lifetime. Consequently, the impact of 241Am buildup was confirmed to be negligible in the safety assessment of geological disposal with an intrinsically large safety margin.Graphical abstractPlots illustrating the effect of spent fuel cooling-off period on the thermal properties of vitrified high-level waste (HLW) and its long-term disposal safety. The left shows the results of waste volume reduction by higher waste loadings in the context of thermal constraint of the bentonite buffer less than 100 °C. The center shows the dissolution rate after disposal depending on the temperature of glass surface. The right is the result of safety assessment for different dissolution rates.

Issues Associated with Yellow Phase Formation and Other Disruptions in Japanese High-Level Waste Vitrification and Their Improvement

The Rokkasho Reprocessing Plant, located in northern Japan, includes facilities for reprocessing spent nuclear fuels and immobilizing high-level waste (HLW) into a stable-glass waste form based on borosilicate. The vitrification process consists of two large liquid-fed, joule-heated, ceramic-lined melters (LFCMs). During the active test campaigns at the Rokkasho Vitrification Facility, unstable melting performance and difficulty in glass operation were observed in the LFCM operation. The operating protocols that had been developed in the earlier full-scale mock-up testing with inactive waste simulants were found to be inadequate to manage a stable LFCM operation. We needed to understand correctly the fundamental causes for the operational difficulties experienced in the active test. First, we studied and summarized the troublesome element behaviors and disturbances in the HLW vitrification process, primarily on the LFCM. Second, we proposed an empirical model, including the transition of the phenomena that appeared in the active LFCM operation. Third, we analyzed the operating data and the trend of operation indexes to find out the fundamental causes for the unstable LFCM operation based on the results obtained in the described model. It was revealed that the presence of a significant amount of molybdenum and sulfur, together with the noble metals (NMs) Ru, Rh, and Pd in the Rokkasho HLW stream could lead to the formation of molybdate and sulfate salt phases [yellow phases (YPs)], and subsequently, the sedimentation of NMs at the bottom of the melter. The YPs may concentrate at the melter bottom and clog the pouring nozzle prior to the bottom accumulation of NM sludge, which can disrupt power deposition and cause difficulties during glass pouring operation. Based on these investigated results, we provide insights into short-term countermeasures, including numerical control of the heat balance in the melter, dilution of HLW feed with inactive waste simulants, and periodic rinsing of the bottom glass. We also provide an evaluation of their validities on the active operation through full-scale mock-up testing. However, the periodic rinsing and dilution operation produce additional glass-filled canisters and increase the total backend cost. Finally, we see a couple of important directions for future research and development. An advanced LFCM system has been developed for the long-term countermeasures to introduce recent advances in the LFCM design with YP and NM compatibilities. Furthermore, the modified or new glass formulations, which can incorporate more waste oxides into the glass matrix, known as high-waste loading glasses, also have been developed in parallel.

Co-immobilization of cesium and strontium containing waste by metakaolin-based geopolymer: Microstructure, mineralogy and mechanical properties

The present study investigated the possibility of co-immobilization of caesium and strontium containing waste by metakaolin-based (MK) geopolymer. 4% of CsNO3 and 4% of Sr(NO3)2 by weight were used as simulants for radioactive Cs and Sr. The effect of Cs and Sr on the geopolymerization kinetics was determined via isothermal calorimetry. Furthermore, the evolution of mechanical properties, microstructure, and mineralogy was analysed. The results showed that Cs exerted no noticeable effect on reaction kinetics while Sr lowered the geopolymerization degree of MK geopolymer due to part of OH− from the activating solution being consumed by Sr2+, thereby reducing the activating capacity of pore solution. In addition, the pore structure analysis showed that the incorporation of Sr coarsened the microstructure of MK geopolymer while no significant change was observed with Cs. The phase analysis showed that no new phase was formed because of Cs or Sr incorporation. The mechanical properties decreased with Cs and Sr addition and a higher strength loss was observed for Sr-bearing geopolymers than that of Cs-bearing matrices. The strength reduction in Cs-bearing matrices is associated with the accompanying nitrate ions while a more pronounced strength reduction in Sr-bearing specimens is due to the synergistic effect of nitrates and Sr2+ on the geopolymerization. Despite that, the compressive strengths for all tested MK recipes met the minimum Belgian waste form acceptance criteria indicating that MK geopolymer preserved sufficient strength even at high Cs and Sr waste loadings. Thermodynamic modelling based on speciation and saturation index revealed that Cs+ was mainly bound through ion exchange, whereas Sr2+ largely precipitated as SrCO3. More importantly, it was observed that the co-immobilization of Cs and Sr by MK geopolymer is largely influenced by Sr behaviour. Based on the findings, it can be concluded that MK geopolymer is an excellent host material for Cs and Sr-based waste.

Effect of Si/Al molar ratio and curing temperatures on the immobilization of radioactive borate waste in metakaolin-based geopolymer waste form

Immobilization of radioactive borate waste (RBW) using a geopolymer with a high Si/Al ratio has been challenging because boron-silicon networks lower the compressive strength and delay the setting time. In this study, metakaolin-based geopolymer waste form to immobilize simulant RBW was fabricated using different Si/Al ratios (1.0–1.4) and curing temperatures (26 and 60 ℃). The 7-day compressive strength results revealed that a certain amount of silicon and an elevated curing temperature are required to achieve high compressive strength and waste loading. Following waste acceptance criteria tests, all geopolymers exhibited compressive strengths higher than 3.445 MPa. The leachability index of boron was higher than 6.0, and the leaching mechanism was identified as diffusion. No significant structural changes in the geopolymer were observed after thermal cycling and gamma irradiation tests. The physically bound or unincorporated RBW was leached out of the geopolymer during water immersion and leaching tests; however, boron, which was chemically connected with silicon, was present as an inert phase together with a geopolymer binder. Consequently, immobilizing RBW using a geopolymer with a low Si/Al ratio (1.4) is beneficial in terms of RBW loading and structural durability.

Dehalogenation reactions between halide salts and phosphate compounds.

Reactions between phosphoric acid [H3PO4] or ammonium hydrogen phosphates [i.e., NH4H2PO4, (NH4)2HPO4] and halide salts can be used to dehalogenate (remove halides from) salt-based waste streams, where the process of removing halides yields products that have more efficient disposal pathways for repository storage. In this context, the term efficiency is defined as higher waste loadings and simplified immobilization processes with potential for recycle of certain salt components (e.g., 37Cl as H37Cl or NH4 37Cl). The main streams identified for these processes are nuclear wastes generated during electrochemical reprocessing of used nuclear fuel as well as used halide salts from molten salt reactor operation. The potential byproducts of these reactions are fairly consistent across the range of halide species (i.e., F, Cl, Br, I) where the most common are hydrogen halides [e.g., HCl(g)] or ammonium halides (e.g., NH4Cl). However, trihalide compounds (e.g., NCl3), nitrogen triiodide ammine adducts [NI3·(NH3) x ], and ammonium triiodide (NH4I3) are also possible. Several of these byproducts (i.e., NCl3, NBr3, NI3, and NH4I3) are shock-sensitive contact explosives so their production in these processes must be tracked and carefully controlled, which includes methods of immediate neutralization upon production such as direct transport to a caustic scrubber for dissolution. Several benefits arise from utilizing H3PO4 as the phosphate additive during dehalogenation reactions for making iron phosphate waste forms including more oxidized iron (higher Fe3+:Fe2+ ratios), higher chemical durabilities, and the avoidance of trihalides, but the byproducts are hydrogen halides, which are corrosive and require special handling.

Cs3Bi2I9-hydroxyapatite composite waste forms for cesium and iodine immobilization

Perovskite-based ceramic composites were developed as potential waste form materials for immobilizing cesium (Cs) and iodine (I) with high waste loadings and chemical durability. The perovskite Cs3Bi2I9 has high Cs (22 wt%) and I (58 wt%) content, and thus can be used as a potential host phase to immobilize these critical radionuclides. In this work, the perovskite Cs3Bi2I9 phase was synthesized by a cost effective solution-based approach, and was embedded into a highly durable hydroxyapatite matrix by spark plasma sintering to form dense ceramic composite waste forms. The chemical durabilities of the monolithic Cs3Bi2I9 and Cs3Bi2I9-hydroxyapatite composite pellets were investigated by static and semi-dynamic leaching tests, respectively. Cs and I are incongruently released from the matrix for both pure Cs3Bi2I9 and composite structures. The normalized Cs release rate is faster than that of I, which can be explained by the difference in the strengths between Cs−I and Bi−I bonds as well as the formation of insoluble micrometer-sized BiOI precipitates. The activation energies of elemental releases based on dissolution and diffusion-controlled mechanisms are determined with significantly higher energy barriers for dissolution from the composite versus that of the monolithic Cs3Bi2I9. The ceramic-based composite waste forms exhibit excellent chemical durabilities and waste loadings, commensurate with the state-of-the-art glass-bonded perovskite composites for I and Cs immobilization.

Production of Feather-Based Biopolymers as a Direct Alternative to Synthetic Plastics

The diminishing supply of crude oil has led the polymer industry to continue to investigate the use of alternative polymer feedstocks. Agricultural waste streams (or low-value byproducts) often contain the building blocks for biopolymer production. One potential feedstock is poultry feathers. Global poultry production currently exceeds 50 billion birds per annum and is continuing to increase. Generating more than 6 million tonnes of feathers/keratin annually, it represents a viable resource for biopolymers. A barrier to industrial adoption has been scalability. This work aims to upscale the extrusion process by adapting to a larger extruder (capable of 3+ kg/h), which will also improve mix stability. Use of this system improved throughput to 0.8 kg/h. The pellets produced were then used in secondary processing (tube and sheet extrusion) to investigate the potential for targeted applications. Analysis on water uptake, morphology, and mechanical properties was conducted to characterize feather-based polymers. Mechanical properties (tensile strength 3.6 MPa and elongation 58%) were similar to those of other biopolymers (range between 0.3 and 10.6 MPa and 2 and 85%, respectively) yet were achieved using a lower plasticizer content (24% compared to 29–40%)/higher waste loading. Results proved that feather polymers could be produced on a large scale and used to produce a range of final applications using a material containing ∼70% waste product.

The effects of mixing multi-component HLW glasses on spinel crystal size

The Hanford Waste Treatment and Immobilization Plant will vitrify radioactive waste into borosilicate glass. The high-level waste (HLW) glass formulations are constrained by processing and property requirements, including restrictions aimed at avoiding detrimental impacts of spinel crystallization in the melter. To understand the impact of glass chemistry on crystallization, two HLW glasses precipitating small (∼5 μm) spinel crystals were individually mixed and melted with a glass that precipitated large (∼45 μm) spinel crystals in ratios of 25, 50, and 75 wt%. The size of spinel crystals in the mixed glasses varied from 5 to 20 μm. Small crystal size was attributed to: (1) high concentrations of nuclei due to the presence of ruthenium oxide and (2) chromium oxide aiding high rates of nucleation. Results from this study indicate that the spinel crystal size can be controlled using chromium oxide and/or noble metal concentrations in the melt, even in complex mixtures like HLW glasses. Smaller crystals tend to settle more slowly than larger crystals, therefore smaller crystals would be more acceptable in the melter without a risk of failure. Allowing higher concentrations of spinel-forming waste components in the waste glass enables glass compositions with higher waste loading, thus increasing plant operational flexibility. An additional benefit to the presence of chromium oxide in the glass composition is the potential for the oxide to protect melter walls against corrosion.

Perovskite-Derived Cs2SnCl6-Silica Composites as Advanced Waste Forms for Chloride Salt Wastes.

Advanced materials and processes are required to separate halides and fission products from complex salt waste streams associated with the chemical reprocessing of used nuclear fuels and molten salt reactor technologies for immobilization into chemically durable waste forms. In this work, we explore an innovative concept using metal halide perovskites as advanced host phases to incorporate Cs and Cl with very high waste loadings. Wet chemistry-synthesized Cs2SnCl6 powders from CsCl salt solutions are successfully encapsulated into a silica matrix to form a composite using low-temperature spark plasma sintering with tunable Cs and Cl loadings up to 31 and 26 wt %, respectively. Chemical durability testing of the composite waste forms by semi-dynamic leaching experiments demonstrates that an incongruent leaching mechanism dominates the release of Cs and Cl. The metal halide perovskite-silica composite waste forms display exceptional chemical durability with the long-term release rates of Cs and Cl comparable to or outperforming the state-of-the-art waste form materials but with significantly higher waste loadings. The scalable synthesis of the metal halide perovskite from wet chemistry processes opens up new opportunities in designing advanced waste forms for salt wastes with very high waste loadings and exceptional chemical durability for the sustainable development of advanced fuel cycles and next-generation reactor technologies.

Immobilization of cesium and iodine into Cs3Bi2I9 perovskite-silica composites and core-shell waste forms with high waste loadings and chemical durability

Cs3Bi2I9, a defect perovskite derivative, is a potential host phase to immobilize iodine and cesium with high waste loadings. In this work, two strategies were explored to form Cs3Bi2I9-silica composites and a core-shell structure in order to improve chemical durability of waste form materials meanwhile maintaining high waste loadings. Cs3Bi2I9 loadings as high as 70 wt.% were incorporated into a silica matrix to form silica-ceramic composites, and 20 wt.% Cs3Bi2I9 was encapsulated into silica to form a core–shell structure by low temperature spark plasma sintering. Chemical durability of the composite and core-shell waste forms was evaluated by semi-dynamic leaching experiments, and Cs and I were incongruently released from waste form matrices. A BiOI alteration layer formed, acting as a passivation layer to reduce the release of radionuclides. The long-term iodine release rate was low (30 mg m−2 day-1) for the 70 wt.% Cs3Bi2I9–silica composite leached in deionized water at 90 °C, which can be further reduced to 5 × 10−3 mg m−2 day−1 for the 20 wt.% core-shell structure. This work highlights a robust way to immobilize the highly mobile radionuclides with high waste loadings through encapsulation into durable matrices and a surface passivating mechanism that can greatly reduce the elemental transport from waste form materials and significantly enhance their chemical durability.

Removal of solids and chemical oxygen demand in poultry litter anaerobic digestion with different inocula

Population growth has contributed to increasing poultry production, entailing a high waste loading, mainly poultry litter. One of the alternatives to treat such residues is anaerobic digestion, in which digester startup and generated-digestate quality are related to the material to be digested and to operation conditions, wherein inoculum use is one of the factors. This study therefore aimed to investigate how digestates, such as inocula, influence poultry litter (PL) anaerobic digestion, as well the reduction of solids and chemical oxygen demand (COD). For this, two inocula (bovine and swine digestates) were tested in the digestion process. The inocula were added at loads of 0.67, 1.00 and 1.67 gVS.L-1day-1. A split-plot design was developed and data underwent analysis of variance with means compared by the Tukey's test at 5% significance. Concerning bovine and swine inocula, it was concluded that both are indicated in the process. However, swine inoculum is better indicated because it had a better removal of total solids (TS), volatile solids (VS) and COD.

Evolution of cations speciation during the initial leaching stage of alkali-borosilicate-glasses

AbstractAlkali-borosilicate glasses (ABS) are used as host immobilization matrices for different radioactive waste streams and are characterized by their ability to incorporate a wide variety of metal oxides with respectively high waste loadings. The vitreous wasteform is also characterized by very good physical and chemical durability. The durability of three ABS compositions were analyzed by investigating their leaching behavior using the MCC1 test protocol and these data were used to investigate the waste components retention in the altered layer and the evolution of the interfacial water composition during the test. The results indicated that the Mg species evolution is exceptional with respect to other alkaline elements and dependent on glass matrix composition and leaching progress, while transition elements speciation is fairly constant throughout leaching process and independent on glass compositions. Si and B species are changing during leaching process and are affected by waste composition. For modified wasteform sample, evolution of Mg, Si and B species is respectively constant, whereas at highest waste loading, these elements have fairly constant speciation evolution within the first 2 weeks of leaching.

Influence of Transition Metal Charge Compensation Species on Phase Assemblage in Zirconolite Ceramics for Pu Immobilisation

ABSTRACTImmobilisation of Pu in a zirconolite matrix (CaZrTi2O7) is a viable pathway to disposition. A-site substitution, in which Pu4+ is accommodated into the Ca2+ site in zirconolite, coupled with sufficient trivalent M3+/Ti4+ substitution (where M3+ = Fe, Al, Cr), has been systematically evaluated using Ce4+ as a structural analogue for Pu4+. A broadly similar phase assemblage of zirconolite-2M and minor perovskite was observed when targeting low levels of Ce incorporation. As the targeted Ce fraction was elevated, secondary phase formation was influenced by choice of M3+ species. Co-incorporation of Ce/Fe resulted in the stabilisation of a minor Ce-containing perovskite phase at high wasteloading, whereas considerable phase segregation was observed for Cr3+ incorporation. The most favourable substitution approach appeared to be achieved with the use of Al3+, as no perovskite or free CeO2 was observed. However, high temperature treatments of Al containing specimens resulted in the formation of a secondary Ce-containing hibonite phase.

Probing swift heavy ion irradiation damage in Nd-doped zirconolite

Synroc minerals have been studied since decades for immobilization of high-level radioactive wastes released from nuclear weapons and nuclear reactors. Among them, the zirconolite, due to its high radiation and aqueous resistance and high waste loading, is considered as a potential candidate for containment purposes. Though irradiation damage effects of swift heavy ions in the material remain a major concern of study, a single phase Nd-doped zirconolite samples prepared through conventional solid-state reaction were irradiated with 120 MeV Au+ ions to study radiation damage effects. Irradiation-induced effects were investigated using ex-situ XRD and micro-Raman spectroscopic techniques. Zirconolite has been observed to undergo amorphization which is confirmed through disappearance of characteristic x-ray diffraction peaks and appearance of broad diffuse scattering bands. Amorphization trend with fluence appeared linear at lower fluences followed by saturation at higher fluences inferring therewith the formation of ion tracks. The ion fluence dependent Raman examination was also found to be in line with the XRD observations. However, the presence of Raman modes, characteristic of zirconolite in the 100–1000 cm−1 region, showed intact behavior of the local structure. To sum up, the 120 MeV heavy Au+ ions developed columnar defects, vacancies and induced micro-strains that lead to the bonds distortion and eventually amorphization of the zirconolite though retaining local ordering.

Thermal footprint of a geological disposal facility containing EURO-GANEX wasteforms

Synroc-Z is a new wasteform being developed to be specifically suited to waste from the advanced EURO-GANEX reprocessing route. This type of waste contains nuclides that release significant amounts of heat during their decay. Here, predictions for radiogenic heat generation from EURO-GANEX wasteforms based on Synroc-Z ceramic and R7T7 glass materials are presented. During reprocessing, any actinides present in the waste-stream are removed to leave only fission products. Simulation results are given demonstrating that, although initially classified as high level waste, the effects of decay mean the wasteforms effectively become intermediate level after 140 years of storage. Furthermore, many of the issues arising from alpha decay will be avoided. During the first 600 years, it is shown that the main source of self heating comes from the decay of Cs-137 and Sr-90. Our calculations show how the significant radiogenic heating by these wasteforms must be considered when defining package sizes and waste loadings. These effects are discussed for both ceramic and glass wasteforms and demonstrate that Synroc-Z provides more flexibility allowing larger packages even at higher waste loadings than R7T7.

Glass ceramic for the vitrification of high level waste with a high molybdenum content

Vitrification has been selected in France as the immobilization process for high-level waste arising from spent fuel reprocessing. Several high-level waste solutions from the reprocessing of legacy UMo spent fuel, used in gas cooled reactors, have been stored in the Orano La Hague facility in stainless steel tanks since the mid-1960s. A special glass-ceramic formulation has been developed and qualified through lab and pilot testing to meet standard waste acceptance criteria for the final disposal of the UMo waste. These solutions are very rich in molybdenum and phosphorus whose contents make the molten glass quite corrosive and require a high-temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum oxide). Molybdenum is known to be sparingly soluble in conventional borosilicate glass. The formulated glass-ceramic matrix comprises a major vitreous phase containing secondary phase particles less than 100 μm in diameter. These are formed by phase separation and crystallization as the molten glass cools. The physical and microstructural properties of the UMo glass in the solid and liquid states are reported here. Evolution of microstructure as a function of the cooling profile was investigated, given the sensitivity of the crystallization process to the cooling profile. The chemical durability of the UMo glass-ceramic was also studied. The feasibility of this process has been demonstrated in a full-scale pilot facility with inactive surrogate solutions.

Development of metakaolin-based geopolymer for solidification of sulfate-rich HyBRID sludge waste

Radioactive waste which contains high sulfate ion is an abiding challenge for cementitious solidification due to the formation of sulfate mineral such as ettringite. Such secondary mineral formation can cause the expansion and disintegration of cement waste form, resulting in the increase of leachability. To alleviate this problem, we proposed metakaolin-based geopolymer waste form to solidify the radioactive waste containing high sulfate content generated from the Hydrazine Based Reductive metal Ion Decontamination (HyBRID) process. Although ettringite formation was observed in the cement waste form, this mineral did not form in the geopolymer waste form. The addition of HyBRID sludge waste significantly increased the compressive strength of geopolymers at Si/Al ratio of 1.6 from 13.6 MPa to 49.6 MPa. The partial dissolution of cristobalite was confirmed during geopolymerization induced by potassium silicate solution, which increases the amount of silicon that enters geopolymerization. The maximum waste loading of 53.8% was achieved when geopolymer composition was K2O∙2.8SiO2∙Al2O3∙15.2H2O, which had compressive strength of 14.3 MPa, exceeding the repository acceptance criterion (3.44 MPa). Potassium-based geopolymer showed high compressive strength, low leachability, and high waste loading. These results increase our understanding of the solidification mechanism for HyBRID sludge waste using geopolymer waste form, and highlight the importance of considering the type of alkali cations and different Si/Al ratios when designing the chemical compositions of geopolymer waste form.

Kinetics of oxyapatite [Ca2Nd8(SiO4)6O2] and powellite [(Ca,Sr,Ba)MoO4] dissolution in glass-ceramic nuclear waste forms in acidic, neutral, and alkaline conditions

Immobilization of chemically complex aqueous waste streams from used nuclear fuel reprocessing is achievable at higher waste loadings with glass ceramics as compared to borosilicate glasses. Additionally, crystalline phases with similar chemistry are more durable than their amorphous counterparts. However, during glass ceramic fabrication, mechanical stresses at crystal-glass interfaces, which are caused by thermal expansion mismatching during cooling, create locales where water is capable of accessing and reacting with the various phases in the glass ceramic, thus releasing radionuclides into the aqueous phase. In the present work, we build on previous chemical durability investigations of a glass-ceramic containing crystalline oxyapatite [Ca2Nd8(SiO4)6O2] and powellite [(Ca,Sr,Ba)MoO4] secondary phases. The individual crystalline and bulk glass phases have been fabricated separately and the corrosion behavior has been investigated with single-pass flow-through (SPFT) testing at 90 °C in buffered pH(RT) 4, 7, 9 and pH 11 solutions and with the static product consistency test (PCT). The results demonstrate the varying dissolution kinetics of the individual phases in the range of pH studies. The consequence of the varying dissolution kinetics are described with a conceptual model of glass-ceramic dissolution mechanisms.

Challenges in the Fabrication of Ceramic Technetium Waste Forms

ABSTRACTIn this research we can demonstrate, that fission technetium-99 can be successfully immobilized as tetravalent cation in solid state refractory oxides such as pyrochlores and perovskites. Pyrochlores show excellent performance in ASTM C1220-10 type corrosion testing and have the ability to structurally bond Tc-99 and therefore avoid the formation of highly-mobile, pertechnetate species under conditions of a generic repository. We have fabricated lanthanide technetium oxides using either dry-chemical ceramic processing, or wet-chemical coprecipitation methods. Tc pyrochlores have shown better Tc retention and corrosion resistance compared with Tc-containing LAWE4-type borosilicate glass, combined with 50-times higher waste loading. However, mechanical properties (fracture toughness, compressive strength) of the pyrochlores are lacking and the microstructure shows high open porosity of about 50 %. To improve these properties we tested a variety of measures such as hot-pressing or the combination of hot pressing and high-temperature synthesis, but the improvement was minor and Tc and the surrogate Ru were partly reduced. The presence of metallic inclusions has strong impact on Tc retention and release rates increased more than tenfold. We have further developed a wet-chemical coprecipitation synthesis route followed by calcination and a 4-days high-temperature sintering cycle for the model composition Sm2(Ru0.5Ti0.5)2O7 where titanium oxide was added as sintering agent. The ceramic surrogate waste forms showed improved theoretical densities of about 73 % combined with sufficient mechanical strength, while maintaining ruthenium in the tetravalent state.

High burn-up operation and MOX burning in LWR; Effects of burn-up and extended cooling period of spent fuel on vitrification and disposal

ABSTRACT Looking ahead to final disposal of high-level radioactive waste arising from further utilization of nuclear energy, the effects of high burn-up of light-water reactors (LWR) with UO2 and MOX fuel and extended cooling period of spent fuel on waste management and disposal were discussed. It was assumed that the waste loading of waste glass is restricted by three factors: heat generation rate, MoO3 content, and platinum group metal content. As a result of evaluation for effects of extended cooling period, the waste loading of waste glass from both UO2 and MOX spent fuel could be increased in the current vitrification technology. For the storage of waste glass from MOX spent fuel with higher waste loading, however, those waste glass require long storage period prior to geological disposal because decay heat of 241Am contributes significantly. Therefore, the evaluation of effects of Am separation on the storage period was performed. Furthermore, heat transfer calculation was carried out in order to evaluate the temperature of buffer material in a geological repository. The results showed, 70 to 90% of Am separation is sufficiently effective in terms of thermal feasibility of a repository.