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.

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.

The low-level nuclear wastes such as decontamination waste from Fukushima are disposed in near-surface underground, where the intermittent recharge of rain and groundwater causes spatial distribution of water content. Therefore, pores of soils are not filled with water, that is, an unsaturated zone will be formed. In such a condition, since the water flow path are detoured by clogged gas in pores of soil in the unsaturated zone, the migration path of radionuclide would be different from the saturated zone. So far, the one-dimensional advection-dispersion equation (ADE) model has been widely used in order to explain experimental results under an unsaturated condition. However, the detouring of local flow-paths remarkably affects the mass transfer. The one-dimensional ADE evaluates such a detouring effect by using Peclet number and retardation coefficient as fitting parameters. In other words, the one-dimensional ADE model is difficult to explain mass transfer under an unsaturated condition. Therefore, the purpose of this study is explaining such complicated transport of radionuclides using a multi-path model based on phenomena in underground. The proposed multi-path model considering both water saturation and permeability distributions showed good agreement with the experimental data under an unsaturated condition.

The US Department of Energy Office of Nuclear Energy is conducting a brine availability heater test to characterize the thermal, mechanical, hydrological and chemical response of salt at elevated temperatures. In the heater test, brines will be collected and analyzed for chemical compositions. In order to support the geochemical modeling of chemical evolutions of the brines during the heater test, we are recalibrating and validating the solubility models for the mineral constituents in salt formations up to 100°C, based on the solubility data in multiple component systems as well as simple systems from literature.

In this work, we systematically compare the model-predicted values based on the various solubility models related to the constituents of salt formations, with the experimental data. As halite is the dominant constituent in salt formations, we first test the halite solubility model in the Na-Mg-Cl dominated brines. We find the existing halite solubility model systematically over-predict the solubility of halite. We recalibrate the halite model, which can reproduce halite solubilities in Na-Mg-Cl dominated brines well.

As gypsum/anhydrite in salt formations controls the sulfate concentrations in associated brines, we test the gypsum solubility model in NaCl solutions up to 5.87 mol•kg–1 from 25°C to 50°C. The testing shows that the current gypsum solubility model reproduces the experimental data well when NaCl concentrations are less than 1 mol•kg–1. However, at NaCl concentrations higher than 1, the model systematically overpredicts the solubility of gypsum.

In the Na+—Cl–—SO42–—CO32– system, the validation tests up to 100°C demonstrate that the model excellently reproduces the experimental data for the solution compositions equilibrated with one single phase such as halite (NaCl) or thenardite (Na2SO4), with deviations equal to, or less than, 1.5 %. The model is much less ideal in reproducing the compositions in equilibrium with the assemblages of halite + thenardite, and of halite + thermonatrite (Na2CO3•H2O), with deviations up to 31 %. The high deviations from the experimental data for the multiple assemblages in this system at elevated temperatures may be attributed to the facts that the database has the Pitzer interaction parameters for Cl–—CO32– and SO42–—CO32– only at 25°C.

In the Na+—Ca2+—SO42–—HCO3– system, the validation tests also demonstrate that the model reproduces the equilibrium compositions for one single phase such as gypsum better than the assemblages of more than one phase.

To estimate the cesium sorption by the bentonite and to obtain the isotherms, some batch-adsorption experiments are being carried out, being the Kd (retardation coefficient) calculated from these isotherms. One-dimensional flow cell was constructed to measure the bentonite permeability regarding to a cesium solution, which results will be used to evaluate the diffusion coefficient – D. It will be used the PHREEQC software to model the transport of the cesium radionuclide through this bentonite with the Kd and D data. The modelling of radionuclide transport in the Brazilian materials will contribute to evaluate the efficiency of multi-barriers system of the national repository, because it is one of its safety criteria.

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.

Can Spent Nuclear Fuel withstand the shocks and vibrations experienced during normal conditions of transport? This question was the motivation for the multi-modal transportation test (MMTT) (Summer 2017), 1/3-scale cask 30 cm drop test (December 2018), and full-scale assembly 30 cm drop tests (June 2019). The full-scale ENSA ENUN 32P cask with 3 surrogate 17x17 PWR assemblies was used in the MMTT. The 1/3-scale cask was a mockup of this cask. The 30 cm drop tests provided the accelerations on the 1/3-scale dummy assemblies. These data were used to design full-scale assembly drop tests with the goal to quantify the strain fuel rods experience inside a cask when dropped from a height of 30 cm. The drop tests were first done with the dummy and then with the surrogate assembly. This paper presents the preliminary results of the tests.

The paper provides generic overview of legal and regulatory framework of radioactive waste management activities held in Russian Federation and national operator responsibilities and accomplishments. It gives a short description of waste classification scheme used and plans for radioactive waste disposal. In particular the paper provides information on the plans of the FEDERAL STATE UNITARY ENTERPRISE "National operator for radioactive waste management" to construct and operate several near-surface disposal facilities for low and intermediate level waste with total capacity up to 550 000 cubic meter. The paper also provides detailed information on the steps of high-level waste disposal program including site-selection, construction phase of the underground research laboratory (URL) near the city of Zheleznogorsk, Krasnoyarsk Region and research program after the construction of the URL. The paper also describes Russian system and state policy in the field of RW management and gives recommendations for future implementers.

Within the framework of the IAEA project “CRAFT” (2011-2014), specialists of JSC “VO “Safety” participated in working group “Safety assessment of the Radon-type facilities”. The IAEA GSG-3 methodology was used in order to address the issue of safety assessment of radioactive waste removal from historical near-surface storage facility of the Radon type. SAFRAN tool (Sweden) was used for safety assessment of a historical Radon type storage facility. Practical experience of SAFRAN application has shown that it can play a significant role in managing records and knowledge on radioactive waste, nuclear facility site, characteristics of geological environment and safety barriers. It can provide reliable long-term storage and effective management of safety related records for the purposes of safety reassessments, review and supervision.