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Pre-compacted MX80 bentonite/Callovo-Oxfordian (COx) claystone mixture has been considered as a promising sealing/backfilling material in deep geological repository for high-level radioactive waste in France. When the pre-compacted blocks are emplaced in the gallery, technological voids can remain. After the infiltration of groundwater from the host rock, they will swell freely, filling the technological voids, and then undergo further hydration under constant-volume conditions. From the perspective of the storage safety, it is essential to understand the effects of technological voids and hydration time on the hydro-mechanical behaviour of such pre-compacted blocks. In this work, a series of infiltration tests at various hydration times was carried out on compacted MX80 bentonite/COx claystone mixture with different technological voids. The evolutions of the swelling pressures in axial and radial directions as well as the hydraulic conductivity were monitored while wetting. After the predetermined hydration times, the dry density, water content and suction at different positions were determined, together with the microstructure investigation using mercury intrusion porosimetry. It was observed that the soils close to initial voids swelled and filled the initial voids upon contact with water, with significant increases in large-pore and medium-pore void ratios. From the variation of dry density profile with time, compression and swelling zones could be identified: in the compression zone, the soils corresponding to the initial soils were subjected to compression, with decrease in large-pore and medium-pore void ratios over time, whereas in the swelling zone, the soils with a higher dry density than the expected final one underwent further swelling, with large-pore void ratio increasing until saturation and then a slight decrease due to water redistribution in the soil. Owing to the soil density heterogeneity, the axial swelling pressure and hydraulic conductivity of samples with voids were slightly larger than those of samples without voids; contrarily, the radial swelling pressure was lower than that of the samples without voids.

The paper is concerned with the issue of achieving the radiological equivalence (the equivalence of radiation risks) of radioactive waste of nuclear reactors and corresponding mass of natural uranium, taking into account the different migration ability of radionuclides in geological formations and soil. This migration radiological equivalence is being investigated for the deep burial of radioactive waste in the case of the development of a two-component nuclear power system with the concurrent use of thermal neutron reactors and fast neutron reactors. Calculations were performed of radiation doses and radiation risks of cancer death arising from consumption of drinking water from a well above a disposal site. The radiation risk relating to a two-component nuclear power system is lower than that from natural uranium; i.e., after reaching the radiological equivalence (100 y of storage) over the timescale of 109 y, the principle of migration radiological equivalence is satisfied. It would take 106 y after radioactive waste disposal to reach the migration radiological equivalence if only thermal reactors were operated. As regards consumption of well drinking water, the radiation risk does not exceed 10-5 y-1 for a two-component nuclear power system, while being 10-3 y-1 (socially unacceptable level) for a power system using only thermal reactors. Radionuclides 241Am, 239Pu, and 240Pu in drinking water make the main contribution to the doses and radiation risks of people for 104 y after the disposal of radioactive waste.

Ensuring the thermal stability of heat-generating nuclear waste glass canisters in interim storage and the thermal stability of bentonite in the deep geological repository are crucial to preserving...

This paper deals with the development and research of low pH concrete mixtures intended for the deep geological repository for radioactive waste. Mixtures, in particular those made from materials originating in the Middle Europe, specifically Czech Republic, were designed due to the diversity of the raw materials’ properties. In this first part, a large number of cement mortars were designed, on which the effect of active ingredients (microsilica, slag) on the pH value over time were tested. From the results of the cement mortars’ pH measured after 90 days, a concrete formula was designed. A mixture with a lower proportion of cement and a higher proportion of microsilica seems to be the best. The slag serves only as a supplement in the place value of several weight percent of the total binder volume. Also, the control measurements of basic material characteristics, such as volumetric mass density and compressive strength were performed on these mixtures.

Canada has decided on long-term containment and isolation of used fuel within a deep geological repository. The deep geological repository will be constructed at a depth of roughly 500 m and consist of a network of placement rooms at depth and surface facilities to support the operation of the repository. An engineered barrier system will fill the area between used fuel containers and the geosphere to isolate the used fuel containers from the environment. The engineered barriers will be subjected to competing gradients of thermal energy driving moisture away from the containers and hydraulic gradients driving moisture into the repository. A key thermal design criterion is for the container surface temperature to be less than 100°C. Recently, thermal properties were defined for the barrier materials in the placement room. The effect of barrier material saturation on dissipation of thermal energy for a generic Canadian placement room is examined for upper- and lower-bound scenarios, which are believed to bound the short-term thermal response of the repository under hydraulically static conditions. The results show that the 100°C criterion is met if continuity is maintained between the container and the rock.

The article presents analysis of safety of long-lived high level radioactive waste (RW) disposed for permanent storage in a deep geological repository. According to the study protocol it is necessary to estimate the human and environmental health safety of RW, generated in different nuclear fuel cy-cles, during their long-term stay in the repository. To estimate biological effects, RW composition from thermal reactors in open nuclear fuel cycle (“TR”) and from simultaneously operated thermal and fast reactors in closed nuclear cycle (“TR+FR”) in different time after their disposal have been evaluated. It is suggested that RW to be safe to humans and the environment if committed effective doses from RW components including minor actinides and fission products and from natural urani-um are equivalent (radiation equivalence) or lifetime attributable risks of cancer mortality from RW and natural uranium are equivalent (radiological equivalence). For this purpose it is important to evaluate the time of achievement of radiation or radiological equivalence. To evaluate health effect of RW containing radio-nuclides on the public and the environment their migration activity has been evaluated with the use of distribution coefficient and retardation factor. Probable annual effective doses to the public from exposure to the radionuclides released from the repository to the biosphere at different time after the RW disposal have been estimated. Radiation doses to the public following the consumption of water from the well installed on the repository, and lifetime attributable risk (LAR) of cancer mortality associated with the exposure to radionuclides contained in “TR” and ”TR+FR” generated waste have been evaluated. It turned out that LARs of cancer mortality from “TR+FR” radionuclides is lower than LAR from natural uranium, it means that radiological migration equivalence after its achievement will exists until the repository conditions allow, about 1 billion years. We have found that LARs from “TR+FR” waste is about 100 times lower than the negligible level, 10-6, established in the Russian National Radiation Safety Standards (NRB-99/2009). Estimat-ed LAR of cancer mortality from “TR”-generated waste is 360 times higher than LAR from “TR+FR”-generated waste. As evaluated for the waste from TR radiological equivalence will be achieved not earlier than 1 million years after permanent disposal. If RW is generated in the closed cycle estimat-ed LAR value from consumption of well water does not exceed 10-5 year-1, if RW is generated in open fuel cycle, the LAR value is 10-3 year-1, this is socially unacceptable risk. In the first 10 thousand years after the disposal the major dose- and risk-forming radionuclides in the well water will be 241Am, 239Pu и 240Pu. Estimates of radiation detriment from ”TR” generated waste put to the perma-nent deep geological disposal are given in the article.

In Germany it is planned to directly dispose spent nuclear fuel (SNF) from nuclear power plants together with other high-level radioactive wastes (HLW) from former SNF reprocessing (e.g., vitrified waste), in a deep geological repository for heat-generating wastes – the siting process for this repository was started in 2017 and is ongoing. Based on several decades of research, development, and demonstration (RD&D) it is generally accepted at the technical and scientific level that direct disposal of HLW and SNF in deep mined geological repositories is the safest and most sustainable option (CEC, 2011; IAEA, 2004). The current efforts to improve the performance and accident tolerance of fuels in nuclear power generation resulted in an increased utilization of a variety of new types of light-water reactor (LWR) fuels such as fuels doped with Cr, Al, and Si. This doping leads to a significant change of the microstructure of the fuel matrix. The corrosion behavior of these types of fuels under conditions relevant to deep geological disposal has hardly been studied so far; however, this is of crucial importance as the development of a robust safety case for deep geological disposal of SNF requires a solid understanding of its dissolution behavior over very long time scales (up to 1 million years). To fill this knowledge gap, additional systematic studies on modern doped UO2 fuels were needed. Corrosion experiments with SNF cannot entirely unravel all of the various concurring effects of the dissolution mechanism due to the chemical and structural complexity of SNF and its high beta and gamma radiation field during the first 1000 years; moreover, technical restrictions only allow a very limited number of experiments. Therefore, within the EU-DisCo project (https://www.disco-h2020.eu, last access: 11 October 2021), a very ambitious programme of corrosion studies on irradiated Cr and Al/Cr doped fuels was carried out, which was complemented by systematic single-effect dissolution studies (e.g., with respect to doping level, grain size and thermodynamic aspects) performed on carefully prepared and characterized, simplified UO2-based model materials. Here, we present recent results on the dissolution behavior of tailor-made UO2 model materials in accelerated static batch experiments using H2O2 as simulant for radiolytic oxidants, present in long-term disposal scenarios for SNF in failed container conditions due to the alpha irradiation of water. In these dissolution experiments pure UO2 reference pellets exhibiting different densities and grain sizes, as well as Cr-doped UO2 pellets with various Cr-doping levels, produced using different doping methods having different grain sizes, were used. In addition, Nd-doped and industrially produced Cr- and Cr/Nd-doped UO2 pellets were used to determine the influence of these parameters on the dissolution rates. The dissolution experiments were performed under strictly controlled conditions with respect to exclusion of oxygen, temperature control, and exclusion of light. This bottom-up approach was followed to understand how the addition of Cr-oxide into the fuel matrix affects SNF dissolution behavior under repository relevant conditions. The results of the dissolution experiments performed with real SNF and the model materials obtained by the DisCo partners build the basis for numerical simulations on the dissolution behavior of modern SNF. First results of the data evaluation indicate that the addition of dopants and the consequential modification of the fuel matrix does not lead to a significant change of the dissolution behavior of these fuels under repository relevant conditions compared to standard SNF (i.e. dissolution rates agree within an order of magnitude).

This paper analyses socially optimal nuclear plant operation and nuclear waste management. Two waiting rules are derived: the first characterizes the optimal continuation of electricity production, and the second gives the optimal nuclear waste disposal date. Both rules balance the cost and benefit of either continuing production or delaying waste disposal into a deep geological repository. In addition, multiple regulatory options are investigated. The optimized waste storage and disposal cost forms the payment that should be collected from the nuclear power firm into a nuclear waste fund. The properties of this payment and other regulatory options including a tax to be paid at the shutdown date of the plant are investigated, and it is shown that the money can be collected by a plant-specific constant fee targeted at firm's profit or output. Numerical illustration shows that waste disposal to a deep geological repository is a cost-minimizing solution only with very low interest rates. For interest rates above one percent it is optimal to store the waste in an on-ground storage facility in perpetuity.

Closing the nuclear fuel cycle and transmuting Minor Actinides (M.As) can be considered as an application of the duty of care principlel principle which says that, “before the final disposal of any waste, any possible chemical and/or physical treatment has to be applied in order to reduce the waste’s toxicity, provided the treatment does not convey unacceptable risks or unacceptable costs”. Forty years of complex research and development has shown that Accelerator Driven Systems could provide a solution to the challenge posed by spent nuclear fuels, by enabling the ability to considerably decrease their radiotoxicity lifetime burden and volume. In particular, a multilateral strategy of treatment of the MAs could be a commendable solution for both the countries phasing out the exploitation of nuclear energy and for those pursuing and developing this exploitation. The pre-industrial assessment of the technical and financial feasibility for industrialization is the next step. This applies to the four R&amp, D and Demonstration building blocks: advanced separation, MAs’ loaded fuel fabrication, dedicated transmuters demonstration (MYRRHA) and provision for MAs’ fuel loaded processing. A global vision of the process leading to a sustainable option is proposed.

In the city of Carlsbad, New Mexico, where a deep geological repository named the Waste Isolation Pilot Plant receives regular shipments of military transuranic waste that will remain radioactive for 10,000 years, a group of local actors has been putting major efforts to extend their attachments to nuclear waste futures. In January 2012, Forbes magazine named Carlsbad ‘The Town that Wants America’s Worst Atomic Waste’. Obviously, linking the verb ‘want’ with ‘worst’ is meant to show how unexpected this desire might be. This paper offers an ethnographical account of activities undertaken by their Nuclear Task Force at a peculiar moment. Articulating the relationship between valuation and desiring, it develops the notion of proactive valuation undertakings to refer to local actors’ attempts to generate new attachments and to maintain existing ones through the succession of a series of mediated and equipped valuation undertakings. Rather than approaching desire for waste with scepticism, it follows what ‘desiring’ actors do and how they do it in ‘enterprise form’ and offers a novel, symmetrical and a more comprehensive approach to examine, not only, the co-production of places and techno political orders, but also the constitution of political and moral values at specific sites.

The behavior of spent nuclear fuel exposed to groundwater is crucial in the safety assessment of a deep geological repository for spent nuclear fuel. For this reason, leaching experiments on spent ...

Radioactive waste is planned to be disposed in a deep geological repository in the Opalinus Clay (OPA) rock formation in Switzerland. Copper coating of the steel disposal canister is considered as an option to ensure complete waste containment for a period of 100’000 years. Sulphide is a potential corroding agent to copper. While background sulphide concentrations in the OPA are very low (ca. 10−11 mol/L), it cannot be ruled out that sulphide reducing bacteria (SRB) thriving at discrete locations within the engineered barrier system (EBS) of the repository could generate significantly higher dissolved sulphide concentrations. The aim of this study is to conservatively evaluate the potential for SRB to generate and sustain elevated sulphide concentrations in the repository near-field by considering specific geochemical bottle-necks, to quantify the maximum sulphide fluxes towards the canister, and to assess their potential to corrode the copper canister coating. To address these objectives reactive transport calculations are performed, and main conceptual and parametric uncertainties of the model are assessed.

In the context of the search for a deep geological repository for high-level radioactive waste from nuclear energy a preliminary waste treatment is repeatedly called into play by partitioning and transmutation (P&T). Proponents of this approach promise that with P&T, the requirements for and the risks posed by a – then still necessary – repository could be significantly reduced. However, such technological promises have to be prospectively, promptly and publicly reasonably verifiable. Partitioning is reprocessing in which, in addition to separating uranium and plutonium from the fission products, other material streams (for example, the minor actinides) are extracted. In transmutation, radionuclides – especially through nuclear fission – are converted into other nuclides. Thus, conversion of the parent nuclides into nuclides with shorter half-lives, lower radiotoxicity, or into stable nuclides could be achieved. For the assessment of P&T, essential aspects are the current degree of maturity of necessary technologies, the requirements for research and development, technological development risks, the basic feasibility and objective, risks of a hypothetical operation of corresponding plants and the possible effects on nuclear waste disposal. More specifically, on the technological side, it is all about development periods, technical security requirements and licensability, proliferation risks and implementation periods. The presentation of the results of some hypothetical P&T scenarios is intended to help to assess the impacts on radioactive waste present in Germany, necessary facilities and operating periods. Thus, pyro-chemical and hydrochemical separation processes, special transuranic fuels based on mixed oxides (MOX) or uranium-free fuel types and critical fast reactors, subcritical (accelerator-driven) reactors, as well as molten salt reactors, are considered. One difficulty is that the multiple recycling of the transuranics changes the fuel composition. Detailed statements about these changes are only possible with complex simulation calculations and their influence on safe reactor operation. So far, this has not happened on an international scale. In the modelling presented here, an attempt was made to represent the restrictions that the reactor design has on the fuel composition more precisely, at least insofar as the element composition of the fuel remains the same for the duration of the scenario. Conclusions presented from the analysis of the hypothetical scenarios affect, among other things, necessary operating periods and the number of plants and changes achieved in the stock of both transuranics and fission products.

Nuclear power is an ideal option for sustainable energy sources from U-235 fission. However, this energy generates long-term radioactive waste such as partially used nuclear fuel (PUNF) during electricity production. This work reviews various technologies to provide viable, sustainable, and long-term solutions for the PUNF storage. They include vitrification, partitioning and transmutation (P&T), pyro-processing, and deep geological repository (DGR). Their benefits and drawbacks are evaluated and compared based on previous studies. How to deal with the public perception of DGR and its impacts on the future of nuclear energy and the business opportunities of nuclear storage technology in the global market are discussed. A perspective of recycling nuclear waste into usable fuel is also elaborated. Our literature survey of 160 published articles (1981–2021) showed that DGR is the most ideal solution for long-term storage of the PUNF, as it provides an ultimate destination in a deep underground that permanently isolates the waste from inhabitants and the environment. Although storing PUNF in a DGR maybe convenient and economical in the short-term, the waste must be stored in a retrievable form so that it can be recycled as a fuel. In the long-term, a complete recycling of used nuclear fuel is the best option. As technological solutions and sound radioactive waste management policy are important for the safe storage of PUNF, stakeholders in the nuclear industry should portray long-term radioactive waste management through viable, feasible, and permanent solutions to waste storage for the sake of public safety and the environment.

A deep geological repository has been considered as one of the most appropriate options for the disposal of high-level radioactive waste (HLW), and it will be constructed in a host rock area at a depth of 500 to 1000 m below the ground surface. The geological repository system is based on the concept of an engineered barrier system, and it consists of a disposal canister with packed spent fuel, buffer material, backfill material, and intact rock. The buffer plays an important role to assure the disposal safety of HLW since it can restrain the release of radionuclides and protect the canister from the inflow of groundwater. Since an increased heat quantity is released from the disposal canister into the surrounding buffer material, the thermal conductivity of the buffer material constitutes a key parameter needed to analyze the entire disposal safety. Therefore, this study presents a thermal conductivity prediction model for compacted bentonite buffer material from Kyungju, which is the only benton...

The proliferation resistance (PR) of Th/U and Th/Pu fuels used in CANada Deuterium Uranium for the deep geological repository was assessed by combining the multiattribute utility analysis proposed by Chirayath et al., 2015 [1] with the transport model of radionuclides in the repository and comparing with that of the used natural U fuel case. It was found that there was no significant advantage for Th/U and Th/Pu fuels from the viewpoint of the PR in the repository. It was also found that the PR values for used nuclear fuels in the repository of Th/U, Th/Pu, and natural U was comparable with those for enrichment and reprocessing facilities in the pressurized water reactor (PWR) nuclear fuel cycle. On the other hand, the PR values considering the transport of radionuclides in the repository were found to be slightly smaller than those without their transport after the used nuclear fuels started dissolving after 1,000 years. Keywords: CANada Deuterium Uranium Fuel Cycle, Deep Geological Repository, Multiattribute Utility Analysis, Proliferation Resistance, Transport Model of Radionuclides in Repository, Used Thorium Fuel Disposal

A deep geological repository is the most sustainable way to dispose of high-level radioactive waste. It consists of a multi-barrier system with waste containing canister, sealing material, and geology that avails a facility to provide the long-term isolation of these harmful wastes. The bentonite or sand bentonite blocks serve as a sealing material in the DGR around the waste canisters to isolate them from the surrounding atmosphere. Buffer surrounding the waste canister experiences variation in the temperature for a long duration due to the continuous reduction in the radioactivity and the temperature emitted by the element with time. The thermal flux is induced on the compacted bentonite when a high-temperature waste canister comes in contact with the bentonite buffer and due to seasonal temperature change resulting from a heat source for a long time. As the temperature inside the deep geological repository varies with a long duration of time, it creates a thermal history on the bentonite buffer and changes its physicochemical and hydromechanical properties affecting its overall performance to serve as a sealant. Past reported studies had mentioned the influence of temperature on the various bentonites, but the effect of thermal history on bentonite efficiency is rarely reported. This study presents a review of existing knowledge about the high-temperature impact on bentonite buffer behavior. It has critically investigated the impact of temperature on the physicochemical, hydromechanical, and mineralogical characteristics of bentonite. Further, the analyses are made on the influence and importance of thermal history to be considered while designing the buffer material from few recently reported studies. The physicochemical and hydromechanical properties such as Atterberg’s limits or swell pressure reduced when subjected to heat. It further reduced when time also came into the picture along with the temperature, i.e., thermal history. Such research will be very helpful and will help better understand the different physicochemical and hydromechanical reactions in a bentonite buffer before its designed values are established as part of the engineering barrier systems in the deep-geological repository.

The use of a deep geological repository (DGR) for the long-term disposal of used nuclear fuel is an approach currently being investigated by several agencies worldwide, including Canada's Nuclear Waste Management Organization (NWMO). Within the DGR, used nuclear fuel will be placed in copper-coated steel containers and surrounded by a bentonite clay buffer. While copper is generally thermodynamically stable, corrosion can occur due to the presence of sulphide under anaerobic conditions. As such, understanding transport of sulphide through the engineered barrier system to the used fuel container is an important consideration in DGR design. In this study, a three-dimensional (3D) model of sulphide transport in a DGR was developed. The numerical model is implemented using COMSOL Multiphysics, a commercial finite element software package. Previous sulphide transport models of the NWMO repository used a simplified one-dimensional system. This work illustrates the importance of 3D modelling to capture non-uniform effects, as results showed locations of maximum sulphide flux are 1.7 times higher than the average flux to the used fuel container.

The loading of the nuclear Spent Fuel Assemblies (SFAs) produced in Switzerland into disposal canisters for their final emplacement into a deep geological repository has to be planned in advance. To limit the associated cost and the operational time of the corresponding surface facility, an optimization should be applied to the aforementioned problem, since its solution cannot be found analytically. To that end, the National Cooperative for the Disposal of Radioactive Waste (Nagra) developed the logistics optimization code “SIMAN”, using the so called Simulated Annealing (SA) method. The objective is to find the best possible loading configuration to minimize the total number of canisters needed. The aims of this study were to finish and analyze the performance of the previous code system and to provide features for extended analysis on the resulting canister loading configurations. Two code revisions were used to simulate different boundary conditions regarding the disposal procedure. Realistic boundary conditions confirmed to add a 10% increase (conservatively) to the total number of disposal canisters needed, compared to the solution for the simplified boundary conditions. In addition, the various reasons and their relative contributions to this increase have been determined and quantified. Finally, an extension to the reference disposal concept has been investigated (so-called “mixed canister”) indicating significant potentials for further decrease of the total number of canisters.

Nagra is currently developing design concepts for canisters for the deep geological disposal of spent fuel and high-level waste. A feasibility evaluation study was done to assess a number of candidate canister designs and materials. The potential canister options were assessed with regard to long-term safety by considering mechanical integrity, environmental damage and potential impact on the geological barrier. Manufacturing feasibility, sealing, inspection as well as potential cost were also assessed. A canister concept, based on a thick-walled forged carbon steel substrate coated with copper, was favourably assessed. As a result, development work was undertaken to refine the copper-coated canister design. Different lid designs and closure configurations were considered, using finite element analysis, to determine the limit load by plastic yielding and buckling and to determine safety margins under the design load. In addition, finite element fracture analyses were performed to determine stress ...

The corrosion behaviour of nuclear fuel waste containers depends on the near-field environmental conditions, which may differ significantly from those in the host rock, especially during the early ...

Copper-coated steel containers are part of the engineered barrier system to permanently store Canadian nuclear fuel waste in a deep geological repository. This work models the dose rates (DRs) at t...

The Czech Republic has a well-developed nuclear programme with two nuclear power plants currently in operation and three operational near-surface repositories serving for the disposal of institutional and operational low- and intermediate-level waste. Spent nuclear fuel will be disposed of in a deep geological repository to be constructed in crystalline or metamorphic rock approximately 500 m below the surface. Czech Ca–Mg bentonites will be used as the buffer and backfilling material. Steel-based materials are currently considered as the reference metallic materials in the Czech disposal canister concept. The preferred design is based on a double-walled canister with an inner shell of stainless steel and an outer shell of carbon-steel. Both parts of the canister must provide mechanical stability; the wall of the carbon-steel canister will be strengthened for corrosion-resistance purposes. The presented work focuses on canister design as confirmed by means of mechanical, thermal and criticality ca...

This paper presents a state-of-the-art analysis of the expected degradation processes of a variety of candidate container materials for the disposal of high-level waste and/or spent nuclear fuel. The work, focusing on the most recent developments, has been performed under the auspices of the Implementing Geological Disposal Technology Platform in the context of an international conference hosted by the Nuclear Waste Management Organisation of Canada (NWMO). The scope of the analysis includes the expected corrosion and environmentally assisted cracking behaviour of copper, carbon steel and titanium in contact with relevant buffer materials (e.g. bentonite, cement) and in conditions expected in an underground disposal facility (long-term anoxic conditions). Considerations relative to the expected evolution of the environmental conditions (especially in the period following backfilling) are also presented. Beyond summarising the current state of knowledge, areas in which opportunities for international collaboration may be present are also highlighted. This paper is part of a supplement on the 6th International Workshop on Long-Term Prediction of Corrosion Damage in Nuclear Waste Systems.

Nowadays manipulation and depositing of high-level radioactive waste has become the most important issue, which needs to be solved. High-level radioactive waste consists mainly of spent fuel elements from nuclear power plants, which cannot be deposited for long time in surface repositories in the same way as it is possible in case of low and medium level radioactive waste. The most effective and safe solution in longer time horizon seems to be deep geological repository of high level waste. In this process of deposition, large amount of specific conditions needs to be taken into account while designing the whole underground complex, because the materials and structures must fulfil all necessary requirements. Then adequate safety will be ensured.

To compare two options for disposal of spent nuclear fuels (SFs) a generic GoldSim model for deep boreholes for disposal (DBD) of SFs was developed. As a desktop study, nuclide release and transport from a DBD after the closure of a repository were roughly evaluated and compared to a KBS-3 type disposal in a less deep geological repository (DGD). An assessment result from the DBD is shown to be remarkable and seems to give a sufficient radiological safety margin, compared to the DGD, even though this study was done in a very straightforward manner. A remarkable sensitivity of the travel lengths involved in the fractured geological media around the DBD to the exposure dose rates are not observed with rather fast and short travel times for non-sorbing nuclides with long-half lives.

Partitioning and transmutation (P&T) technologies have been developed for minor actinides (MA) to reduce the high level waste (HLW) volume and long-term radiotoxicity. Although the MA P&T can reduce the potential radiotoxicity effectively by 1-3 orders of magnitude, the actual operation of P&T requires several tens of years for developing elemental technologies of nuclide separation, MA containing fuel fabrication, transmutation and their practical systematization. The high level liquid waste (HLLW) containing MA is presently vitrified immediately after spent fuel reprocessing, stored about 50 years at surface facility and will be disposed of at deep geological repository. Vitrified HLW form works as an excellent artificial barrier against nuclides release during storage and disposal. On the other hand, it is difficult to recover MA from the form. So the present waste management scheme has an issue of MA P&T technology application until its deployment, which will produce much amount of vitrified HLW including long-lived MA without P&T application. Thus the authors proposed the flexible waste management method to increase the effectiveness of the MA P&T. The system adopts the HLLW calcination instead of the vitrification to produce granule for its dry storage of about 50 years until the MA P&T technology will be applicable. The granule should be easily dissolved by the nitric acid solution to apply the typical aqueous MA partitioning technologies to be developed. This paper reports the purpose of the study, the feasibility evaluation results for the calcined granule storage and the evaluation results for the environmental burden reduction effect.

There are two RBMK-1500 type reactors in Lithuania, which are under decommissioning now. Long-lived low and intermediate level radioactive metallic waste generated during operation and decommissioning of the nuclear power plant is foreseen for deep geological disposal (DGR). A preliminary assessment of the radionuclide release from metallic waste and migration through the engineered barriers of the generic DGR in crystalline rocks is presented in this study. The radionuclide flux to backfill and geosphere is investigated for three different container alternatives: (1) disposal of metallic waste in stainless steel containers, (2) disposal in carbon steel containers and (3) disposal in which the container is not a barrier for radionuclide migration. In addition, the impact of surface contamination of the activated waste on the radionuclide flux is investigated. Comparison of maximum fractional flux to backfill between Alternatives 1 and 2 revealed that the container material plays a noticeable role in the migration of long-lived low mobility radionuclides. The maximum fractional flux to the geosphere for all radionuclides increases from Alternative 1 to Alternative 3. Surface contamination is an important parameter when the maximum radionuclide flux to backfill is observed during the early post-closure period. This work was supported by the European Union’s Horizon 2020 Project Theramin (H2020-755480, 2017-2020) and by the Lithuanian state budget under a long-term program (2017-2021).

The near field of the planned deep geological repository for low- and intermediate-level radioactive waste (L/ILW) in Switzerland will consist of different waste materials with different reactivities that are conditioned in cementitious matrices. Geochemical modelling was applied to predict the temporal evolution of the chemical conditions in a cement-stabilised model waste form by simulating the degradation of the waste materials and the alteration of the cementitious matrix. The model waste form contains large amounts of metallic waste and a low amount of organic waste. The geochemical modelling considered the use of either siliceous or calcareous aggregates for fabrication of the cementitious matrix used to condition the waste form. The results show that the type of aggregates has a major effect on the temporal evolution of the chemical conditions in the waste form. The use of calcareous aggregate instead of the commonly used siliceous aggregate is expected to maintain highly alkaline conditions of the waste form over the period of concern for the L/ILW repository. The latter conditions prevent accelerated iron corrosion and, related to that, accelerated H2 production from occurring due to the high pH. The evolution of the waste form is decisively controlled by the availability of water. In the absence of water ingress from the near field, the degradation processes cease as free water in the waste package is exhausted. In the case of saturation achieved by water ingress with time, the degradation processes continue over the entire period of concern for the L/ILW repository. The present study aims to illustrate the potential of geochemical modelling for predicting the temporal evolution of the chemical conditions of a cement-stabilised waste form.

The Nuclear Waste Management Organization is evaluating the safety and feasibility of the permanent disposal of used nuclear fuel in a deep geological repository. Their current design concept utilises copper-coated steel used fuel containers to isolate the waste from the environment. Immediately following repository closure, a finite quantity of O2 will be trapped inside the repository and could cause some amount of oxic corrosion to the outer copper layer of the containers. On a per container basis, 13 mol of O2 will be trapped in the repository rooms at the time of closure, based on reference design dimensions. This corresponds to a maximum depth of copper corrosion of 81 μm, assuming a uniform distribution. This work also considers the sensitivity of this oxic corrosion allowance to various hypothetical design changes to the repository that may occur before or during construction.

The United Arab Emirates is presently formulating a nuclear waste management policy, an element of which relates to the management of spent nuclear fuels, and therefore an analytical assessment of published options is considered timely. This paper considers a suitable assessment methodology and presents a preliminary analysis of four options: reprocessing overseas, take-back by supplier, deep geological repository and deep boreholes disposal. Based on the analysis it is concluded that deep boreholes disposal is the most favourable option for the country. However, all four options have pros and cons and the analysis is sensitive to the sample sets and the sampling group. These issues are discussed further in the paper.

In Slovakia, a direct disposal of spent nuclear fuel in a deep geological repository within the country after a certain period of interim storage is a preferred option. This paper briefly describes near field model of radionuclide migration developed in GoldSim simulation code environment and analyses the calculated results on time-dependent release rates of safety relevant radionuclides. Given the fact that GoldSimalso enables to perform probabilistic simulations using the Monte Carlo method, a probabilistic approach was chosen to assess the influence of selected near field parameter uncertainties related to radionuclide migration on the radionuclide release rates from the bentonite buffer to the surrounding host rock. Based on the results, release rates of nuclides which exceed their solubility limits are effectively lowered and many of nuclides are significantly sorbed on the buffer material. It can be seen that the variance of the total release rate in the case of solubility uncertainty is almost two orders of magnitude within a long period of time.

Recent incidents at the nation's only operating deep geologic nuclear waste repository, the Waste Isolation Pilot Plant (WIPP), resulted in the release of americium and plutonium from one or more defense-related transuranic (TRU) waste containers into the environment. WIPP is a U.S. Department of Energy mined geologic repository that has been in operation since March, 1999. Over 85,000 m3 of waste in various vented payload containers have been emplaced in the repository. The primary radionuclides within the disposed waste are 239+240Pu and 241Am, which account for more than 99% of the total TRU radioactivity disposed and scheduled for disposal in the repository. For the first time in its 15 years of operation, there was an airborne radiation release from the WIPP at approximately 11:30 PM Mountain Standard Time (MST) on Friday, February 14, 2014. The radiation release was likely caused by a chemical reaction inside a TRU waste drum that contained nitrate salts and organic sorbent materials. In a recent news release, DOE announced that photos taken of the waste underground showed evidence of heat and gas pressure resulting in a deformed lid, in material expelled through that deformation, and in melted plastic and rubber and polyethylene in the vicinity of that drum. Recent entries into underground Panel 7 have confirmed that at least one waste drum containing a nitrate salt bearing waste stream from Los Alamos National Laboratory was breached underground and was the most likely source of the release. Further investigation is underway to determine if other containers contributed to the release. Air monitoring across the WIPP site intensified following the first reports of radiation detection underground to ascertain whether or not there were releases to the ground surface. Independent analytical results of air filters from sampling stations on and near the WIPP facility have been released by us at the Carlsbad Environmental Monitoring & Research Center and confirmed trace amounts of 241Am and 239+240Pu, at ratios reflecting the suspect waste stream. The highest activity detected offsite was 115.2 μBq/m3 for 241Am and 10.2 μBq/m3 for 239+240 Pu. These concentrations in air were very small, localized, and below any level of public health or environmental concern.

Spent fuels to be disposed of have various initial 235 U enrichments, discharge burnups, and cooling times, and consequently each type of spent fuel has a different decay heat. Because a disposal system design considering respective spent fuel characteristics was not previously possible in Korea, a system was developed based on a conservative reference spent fuel nominated through historical data analyses on the relation of spent fuel inventory with the initial 235 U enrichment, discharge burnup, and cooling time. Recently, ASOURCE, an advanced source-term calculation system, was developed by the Korea Atomic Energy Research Institute, making it possible to design a disposal system considering the respective source terms of the spent fuels. In this study, the benefits resulting from a consideration of the individual decay heat for entire spent fuels were evaluated in terms of the footprint of a repository. As a result, it was seen that conservativeness in the disposal system design can be remarkably avoided, revealing a roughly 50% decrease in the footprint of the repository, considering the respective irradiation and cooling profile. It is therefore concluded that optimization of the source term is a viable option in a deep geological disposal system design.

The high potential of bentonites to volume changes depending on the water content is considered as their advantage for the engineered barriers in the deep geological repository of high-level radioactive waste and spent nuclear fuel because of swelling and self-healing of cracks in contact with water. On the other hand, drying may lead to opening of cracks and spaces between the bentonite blocks. This would increase the permeability and contamination risk around the hot container with high-level radioactive waste and spent nuclear fuel, especially if the host rock mass is dry. First shrinkage tests on four Slovak bentonites studied for engineered barriers were carried out. The water content at the shrinkage limit and the relative linear shrinkage are the first available shrinkage parameters received for the bentonite paste. The shrinkage hazard is higher in the best bentonites with high swelling potential—from Kopernica and Jelsový potok. The results indicated the necessity of further shrinkage tests to determine the relative linear and volume shrinkage of bentonite elements pressed of the loose bentonite powder of low water content.

Opalinus Clay has been selected as a host rock for a future high-level waste deep geological repository in Switzerland, whereas concepts for spent fuel and high-level waste disposal canisters are currently under development. The development of disposal canisters is done with respect to the expected evolution of the near-field conditions in the repository and requirements set by long-term safety, the handling concept, and manufacturing. A wide range of issues such as manufacturing, materials options, corrosion, closing weld, postweld heat treatment, structural performance, and inspection are considered. A detailed design study has shown that a carbon steel concept may be suitable for a geological repository in Opalinus Clay. Additionally, there are good prospects for using carbon steel design concepts as a substrate for copper coatings, which would largely eliminate gas production as an issue in long-term safety and would avoid the need of postweld heat treatment of steel. In parallel, several possible materials and design options are explored and their feasibility is considered through risk-assessment criteria.

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.

Many countries worldwide are investigating the use of advanced fuels and fuel cycles for purposes such as increasing the sustainability of the nuclear fuel cycle, or decreasing the radiological impact of used fuel. One common metric used to assess the radiological impact to humans of fuels placed in a repository is the total radiotoxicity of the fuel, but this approach does not take into account how engineered and natural (i.e., rock) barriers can remove many radiotoxic nuclides from ground water before they reach the surface. In this study, we evaluate the potential radiological dose consequences of advanced fuels in the context of a full system model simulation for release and transport from a repository, transport through the surrounding geosphere, release to the biosphere and dose consequences for the target critical group. Heavy water moderated reactors, such as the CANDU® reactor, are well-suited to the use of advanced fuels, and the post-closure performance of a deep geological repository for spent natural uranium fuel from them has already been studied. For this study, two advanced fuels of current interest were chosen: a TRUMOX fuel designed to recycle plutonium and minor actinides and thereby reduce the amount of these materials going into disposal, and a plutonium thorium-based fuel whose main goal is to increase sustainability by reducing uranium consumption. The impact of filling a deep geological repository, of identical design to that for natural uranium, with used fuel from these fuel cycles was analyzed. It was found that the two advanced fuels analyzed had dose rates, to a hypothetical critical group of humans living above the repository, which remained a factor of 170 to 340 lower than the current acceptance limit for releases, while being 5.3 (for TRUMOX) and 2.6 (for thorium-plutonium) times higher than those of natural uranium. When the dose rates are normalized to total energy produced, the repository emissions are comparable. In this case, the maximum dose rates were found to be 6% lower for the TRUMOX fuel, and 16% higher for the plutonium thorium fuel, than for the used natural uranium fuel.

A deep geological repository involving a multi-barrier system constitutes one of the most promising options to isolate high-level radioactive waste from the human environment. In order to certify the efficiency of waste isolation, it is essential to understand the behaviour of the confining geomaterials under a variety of environmental conditions. The efficiency of an Engineered Barrier System (EBS) is largely based on thecomplex behaviour of bentonite. To contribute to a better understanding of the processes involved in the EBS, a case study for sensitivity analysis has been defined and is studied using a thermo-hydro-mechanical (THM) finite element approach including a consistent thermo-plastic constitutive model for unsaturated soils. The model also features a coupled THM approach of the water retention curve. Various couplings were studied separately and in combination in order to determine the significance of each. The same principle is applied to physical phenomena such as vapour diffusion. This study clearly highlights the effects that need to be taken into consideration for a correct assessment of EBS behaviour.

This study predicted temperature in the disposal tunnels using computational fluid dynamics based on natural ventilation quantity that comes from high altitude and temperature differences that are the characteristics of high level waste repository. The result of the previous study that evaluated quantitatively natural ventilation quantity using a hydrostatic method and CFD shows that significant natural ventilation quantity is generated. From the result, this study performed the prediction of temperature in disposal tunnels by natural ventilation quantity by the caloric values of the wastes, at both deep geological repository and surface repository. The result of analysis shows that deep geological repository is effective for thermal control in the disposal tunnels due to heat transfer to rock and the generation of sufficient natural ventilation quantity, while surface repository was detrimental to thermal control, because surface repository was strongly affected by external temperature, and could not generate sufficient natural ventilation quantity. Moreover, this study found that in the case of deep geological repository with a depth of 200 m, the heatof about 10℃ was transferred to the depth of 500 m. Thus, it is considered that if the high level waste repository scheduled to be built in the country is designed placing an emphasis on thermal control, deep geological repository rather than surface repository is more appropriate.