Spent Nuclear Fuel Repository Research Articles

UHPC Viability for Nuclear Storage Facilities: Synthesis and Critical Review of Durability, Thermal, and Nuclear Properties for Informed Mix Modifications.

Spent nuclear fuel (SNF) from the United States' nuclear power plants has been placed in dry cask storage systems since the 1980s. Due to the lack of a clear path for permanent geological repository for SNF, consolidated and long-term storage solutions that use durable concrete and avoid current aging and licensing challenges are becoming indispensable. Ultra-high-performance concrete (UHPC) is a rapidly growing advanced concrete solution with superior mechanical and durability properties that can help realize future resilient nuclear storage facilities. Thus, the overall goal of this review study is to demonstrate the viability of UHPC as a long-term solution for future nuclear storage facilities. The paper first identifies all possible non-nuclear (environmental) and nuclear (thermal and radiation-induced) degradation mechanisms in concrete overpacks and storage modules with critical assessment and projections on UHPC performance in comparison to current conventional concrete solutions. Next, since concrete serves as a shielding material in nuclear settings, the preliminary attenuation properties of UHPC from emerging studies are synthesized along with the possible mix modifications to improve its attenuation performance. The paper identifies the major knowledge gaps to inform future research and development, aimed at rethinking the design of SNF dry storage facilities by incorporating UHPC.

Temperature-dependent microbial reactions by indigenous microbes in bentonite under Fe(III)- and sulfate-reducing conditions

Bentonite is a promising buffer material for constructing spent nuclear fuel (SNF) repositories. However, indigenous microbes in bentonite can be introduced to the repository and subsequent sealing of the repository develops anoxic conditions over time which may stimulate fermentation and anaerobic respiration, possibly affecting bentonite structure and SNF repository stability. Moreover, the microbial activity in the bentonite can be impacted by the heat generated from radionuclides decay. Therefore, to investigate the temperature effect on microbial activities in bentonite, we created microcosms with WRK bentonil (a commercial bentonite) using lactate as the electron donor, and sulfate and/or ferrihydrite (Fe(III)) as electron acceptors with incubation at 18 ℃ and 50 ℃. Indigenous WRK microbes reduced sulfate and Fe(III) at both temperatures but with different rates and extents. Lactate was metabolized to acetate at both temperatures, but only to propionate at 18 ℃ during early-stage microbial fermentation. More Fe(III)-reduction at 18 ℃ but more sulfate-reduction at 50 ℃ was observed. Thermophilic and/or metabolically flexible microbes were involved in both fermentation and Fe(III)/sulfate reduction. Our findings illustrate the necessity of considering the influence of temperature on microbial activities when employing bentonite as an engineered buffer material in construction of SNF repository barriers.

Double-faced effects of fission product alloy particles on the behavior of spent nuclear fuel

The purpose of the present work is to investigate the environmental behaviors of spent nuclear fuel (SNF) under deep repository conditions. Over long time periods, oxygen is expected to be introduced into the repository and hydrogen could be generated from an interaction between the iron canister materials and intruding groundwater in future scenarios. The influence of fission product noble metal alloy particles on the stability of the SNF UO2 matrix in a O2 or H2 containing atmosphere were demonstrated by studying the electrochemical and leaching behaviors of UO2 pellets both with and without alloy particle doping. The results show that the doped noble metal alloy particles in UO2 display double faced effects, i.e., can both enhance UO2 oxidative dissolution in an Ar purged, low oxygen containing solution or inhibit the same reaction in the same solution but purged with 4%H2-Ar gas mixture. These conclusions were also verified by the analysis of morphology and composition on the surface of several samples after potentiostatic polarization. The results of this work strongly support the application of a large quantity of iron as the disposal canister inner support material in SNF repository.

On the Nature of the Cathodic Reaction during Corrosion of Copper in Anoxic Sulfide Solutions

Sulfide-induced corrosion is expected to be the dominating long-term corrosion process for copper containers in technical concepts for deep geological disposal of spent nuclear fuel (SNF), adapted in several waste management programs around the world. The present study investigates the atomic-scale mechanism of the cathode side of the corrosion reaction using Density Functional Theory (DFT) calculations. Despite the central role of the reaction, neither the site of reaction nor the active species has been previously established. Here we compare the cathodic reaction leading to H2-evolution on pure copper and on chalcocite (Cu2S) surfaces. The considered H-donors are OH−/H2O and HS−/H2S which are all available at the neutral to alkaline conditions anticipated at the SNF disposal sites. Assuming Volmer-Tafel-Heyrovsky kinetics, we find that the cathodic reactions are many orders of magnitude faster on copper compared to copper sulfide. Although we find that HS−/H2S have lower reaction barriers than H2O, our kinetic analysis suggest that H2O is expected to be the main H-source for the cathodic reaction under SNF repository conditions as results of the low sulfide concentrations (≲ 10 μM) expected in SNF repositories in Sweden, Finland and Canada.

An Analysis of the Factors of Local Acceptance for Spent Nuclear Fuel Repository

An Analysis of the Factors of Local Acceptance for Spent Nuclear Fuel Repository

Evaluation of Uncertainties Related to the Near Field Model of Geological Repository for Spent Nuclear Fuel

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.

Can Shale Safely Host U.S. Nuclear Waste?

Even as cleanup efforts after Japan's Fukushima disaster offer a stark reminder of the spent nuclear fuel (SNF) stored at nuclear plants worldwide, the decision in 2009 to scrap Yucca Mountain as a permanent disposal site has dimmed hope for a repository for SNF and other high‐level nuclear waste (HLW) in the United States anytime soon. About 70,000 metric tons of SNF are now in pool or dry cask storage at 75 sites across the United States [Government Accountability Office, 2012], and uncertainty about its fate is hobbling future development of nuclear power, increasing costs for utilities, and creating a liability for American taxpayers [Blue Ribbon Commission on America's Nuclear Future, 2012].

Ecohydrological Responses to Diversion of Groundwater: Case Study of a Deep-Rock Repository for Spent Nuclear Fuel in Sweden

Planning and license applications concerning groundwater diversion in areas containing water-dependent or water-favored habitats must take into account both hydrological effects and associated ecological consequences. There is at present no established methodology to assess such ecohydrological responses. Thus, this paper describes a new stepwise methodology to assess ecohydrological responses to groundwater diversion from, e.g., water-drained pits, shafts, tunnels, and caverns in rock below the groundwater table. The methodology is illustrated using the planned deep-rock repository for spent nuclear fuel at Forsmark in central Sweden as a case study, offering access to a unique hydrological and ecological dataset. The case study demonstrates that results of ecohydrological assessments can provide useful inputs to planning of monitoring programs and mitigation measures in infrastructure projects. As a result of the assessment, artificial water supply to wetlands is planned in order to preserve biological diversity, nature values, and vulnerable species.Electronic supplementary materialThe online version of this article (doi:10.1007/s13280-013-0404-8) contains supplementary material, which is available to authorized users.

Immobilization of uranium in biofilm microorganisms exposed to groundwater seeps over granitic rock tunnel walls in Olkiluoto, Finland

In an underground rock characterization facility, the ONKALO tunnel in Finland, massive 5–10-mm thick biofilms were observed attached to tunnel walls where groundwater was seeping from bedrock fractures at a depth of 70m. In laboratory experiments performed in a flow cell with detached biofilms to study the effect of uranium on the biofilm, uranium was added to the circulating groundwater (CGW) obtained from the fracture feeding the biofilm. The final uranium concentration in the CGW was adjusted to 4.25×10−5M, in the range expected from a leaking spent nuclear fuel (SNF) canister in a future underground repository. The effects were investigated using microelectrodes to measure pH and Eh, time-resolved laser fluorescence spectroscopy (TRLFS), energy-filtered transmission electron microscopy (EF-TEM), and electron energy-loss spectroscopy (EELS) studies and thermodynamic calculations were utilized as well. The results indicated that the studied biofilms constituted their own microenvironments, which differed significantly from that of the CGW. A pH of 5.37 was recorded inside the biofilm, approximately 3.5units lower than the pH observed in the CGW, due to sulfide oxidation to sulfuric acid in the biofilm. Similarly, the Eh of +73mV inside the biofilm was approximately 420mV lower than the Eh measured in the CGW. Adding uranium increased the pH in the biofilm to 7.27 and reduced the Eh to −164mV. The changes of Eh and pH influenced the bioavailability of uranium, since microbial metabolic processes are sensitive to metals and their speciation. EF-TEM investigations indicated that uranium in the biofilm was immobilized intracellularly in microorganisms by the formation of metabolically mediated uranyl phosphate, similar to needle-shaped autunite (Ca[UO2]2[PO4]2·2–6H2O) or meta-autunite (Ca[UO2]2[PO4]2·10–12H2O). In contrast, TRLFS studies of the contaminated CGW identified aqueous uranium carbonate species, likely (Ca2UO2[CO3]3), formed due to the high concentration of carbonate in the CGW. The results agreed with thermodynamic calculations of the theoretically predominant field of uranium species, formed in the uranium-contaminated CGW at the measured geochemical parameters.This investigation clearly demonstrated that biological systems must be considered as a part of natural systems that can significantly influence radionuclide behavior. The results improve our understanding of the mechanisms of biofilm response to radionuclides in relation to safety assessments of SNF repositories.

Sorption of Np by UO2 under repository conditions

This work is a part of the joint Russian – American Program on Beneficial Use of Depleted Uranium. The production of nuclear fuels results in the accumulation of large quantities of depleted uranium (DU) in the form of uranium hexafluoride (UF6), which is converted to uranium oxides. Depleted uranium dioxide (DUO2) can be used as a component of radiation shielding and as an absorbent for migrating radionuclides that may emerge from casks containing spent nuclear fuel (SNF) that are stored for hundreds of thousands of years in high-level wastes (HLW) and SNF repositories (e.g. Yucca Mountain Project). In this case DU oxides serve as an additional engineered chemical barrier. It is known that the primary radioisotope contributor to the calculated long-term radiation dose to the public at the Yucca Mountain SNF repository site boundary is neptunium-237 (237Np). This paper describes the sorption of 237Np in various media (deionized water and J-13 solution) by DUO2. Samples of DUO2 used in this work originated from the treatment of UF6 in a reducing media to form UO2(DUO2-1 at 600°C, DUO2-2 at 700°C, and DUO2-3 at 800°C). All species of DUO2 sorb Np(V) and Np(IV) from aqueous media. Equilibrium was achieved in 24 hours for Np(V) and in 2 hours for Np(IV). Np(V) sorption is accompanied with partial reduction of Np(V) to Np(IV) and vice versa. The sorption of Np(V) onto DUO2 surfaces is irreversible. The investigations on DUO2 transformations were performed under dynamic and static conditions. Under static conditions the solubility of the DUO2 samples in J-13 solution is considerably higher than in DW. When the pre-treatment temperature is decreased, the solubility of DUO2 samples raises regardless of the media. The experiments on interaction between DUO2 and aqueous media (DW and J-13 solution) under dynamic conditions demonstrated that during 30-40 days the penetration/filtration rate of DW and J-13 solution through a thin DUO2 layer decreased dramatically, and then slowed and stabilized. The filtration rate of J-13 solution through DUO2layer is several times lower than that of DW. If Np were sorbed by UO2 in SNF rather than being transported to the site boundary, the site boundary dose would be more than an order of magnitude lower than that calculated. If this is so, the remaining primary contributors to site boundary dose would be 129I and the decay products of 238U. One could provide an additional barrier, a chemical barrier, to radionuclide release by surrounding the SNF with UO2. The experiments also indicate that a relatively thin layer of UO2 essentially stops water penetration. Such a barrier would deny water access to the SNF. Such a repository SNF configuration would provide reduced radionuclide transport from the drift as well as sorbing released nuclides from failed SNF pins.

The rationality of acceptance in a nuclear community: analysing residents' opinions on the expansion of the SNF repository in the municipality of Eurajoki, Finland

The project to build a final disposal repository for Spent Nuclear Fuel (SNF) in the municipality of Eurajoki, Finland, is approaching its 2012 deadline for the application of a construction licence. At the same time, the nuclear waste company Posiva is already planning to expand the disposal capacity of the repository. This paper addresses the question of acceptance among the residents of Eurajoki regarding the repository's expansion, and examines what aspects should be taken into consideration when explaining local opinions. The local acceptance figures for Eurajoki are analysed in relation to assumptions of the six common explanation types. The relationship between information deficit, social trust, benefit-cost calculation, perceived moral responsibility, perceived risks and threats, and acceptance of nuclear power, on the one hand, and local acceptance of expanding the repository both for current actors and for possible other domestic operators, on the other, is explored. The data provided in the paper is based on a survey carried out in June 2008. The respondents were selected from the residents of the municipality of Eurajoki and the neighbouring municipalities using stratified random sampling (N = 3000). The response rate of the survey was 20% (N = 606).

Uranium Behavior in the Process of Primary Pitchblende Ores Alteration by the Post‐ore Hydrothermal Solutions: An Application to Assessment of Uranium Migration from Underground Spent Nuclear Fuel Repositories

AbstractIt has been shown that the main uranium ore mineral, pitchblende (uranium dioxide), is a natural analog of synthetic uraninite (also uranium dioxide), which constitutes 96% of spent nuclear fuel (SNF). Geochronological studies of the U‐Pb isotope systems in unaltered pitchblende from the orebodies reveal that these systems remained completely closed over the entire period (approximately 135 Ma) since the formation of the deposits. The bulk of the primary uranium ores within the Streltsovskoye ore field was influenced to various degrees by post‐ore hydrothermal solutions that led to pitchblende spherulites being replaced by pseudomorphs of an amorphous phase with a U‐Si composition; this phase also re‐precipitated in veinlets proximal to the pitchblende pseudomorphs. A technique specially developed by the authors was used to carry out quantitative counts of the abundance of uranium minerals by calculating the uranium mass balance in one of the orebodies subjected to hydrothermal alteration. The calculations reveal minimal uranium loss from the orebody. Uranium liberated in the process of the pseudomorphic replacement of pitchblende was immediately fixed, in situ, in the newly formed coffinite‐like amorphous U‐Si phase as a result of the development of an efficient geochemical barrier that prevented the long‐distance migration of uranium. In assessing the long‐term safety of underground SNF repositories, the results of the present study give us confidence that SNF uraninite, in terms of the preservation of its integrity as a mineral phase, provides for the reliable long‐term isolation of uranium, transuranium elements, and fission products that are “sealed” in the uraninite matrix. In the case of the mineral transformation of the uraninite matrix by hydrothermal solutions, the liberated uranium would be efficiently immobilized by the newly formed amorphous U‐Si phase.

Optimization of Spent Nuclear Fuel Filling in Canisters for Deep Repository

We demonstrated the use of combinatorial methods to optimize the filling of spent nuclear fuel (SNF) in metal canisters for final deep SNF repository, according to the maximal allowed thermal power per canister Pmax and the limit of n = 4 spent-fuel assemblies per canister. As a next step, the deposition time can be optimized by minimizing the required number of canisters M and the interim storage time. The method has been tested in detail for a typical pressurized water reactor (PWR), nuclear power plant (NPP) Krsko, SNF for different numbers of reactor cycles and different Pmax. The results show that the time interval between the last reactor cycle and the optimal deposition time varies between 3 and 30 yr for a typical PWR. The most significant contribution to the uncertainty of the calculated SNF decay heat (thermal power) is due to inaccurate cross sections taken from generic cross-section libraries. The quality of the results was verified by comparing the calculated M to the theoretical lower boundary Mmin. The idea behind the optimization method is universal and thus can be implemented for any SNF, canister, and repository design.

The Antei uranium deposit: A natural analogue of an SNF repository and an underground geodynamic laboratory in granite

The estimation of the long-term stability of crystalline rock massifs with respect to natural and technogenic loads in the course of long-term storage of spent nuclear fuel (SNF) is a special area of surveys at underground research laboratories (URLs). In parallel with these surveys, data on uranium deposits—natural analogues of repositories of SNF consisting of 95% UO2—are used for obtaining insight into the dynamics of radionuclide migration and validating barrier properties of host rocks. Examples of URLs located in granitic massifs of Sweden (Aspo), Canada (Whiteshell), Switzerland (Grimsel), Japan (Mizunami), and Finland (ONKALO), as well as the El Berrocal (Spain), Palmottu (Finland), Sanerliu (China), and Kamaishi (Japan) deposits, are considered in the paper. The objects listed above are distinct in tectonic settings, geology, control of ore mineralization, redox conditions of uranium migration, and character and intensity of filtration and transportation, which predetermine the direction and specific features of research conducted therein. A variant in which a URL and a natural analogue are combined in one object is especially promising for validation of safe long-term isolation of SNF. The Antei vein-stockwork uranium deposit in the southeastern Transbaikal region, localized in Paleozoic granite at a depth of 400–1000 m and opened by mine workings at six levels, is such an object. Its geological features, stress-strain state, and infrastructure of mine workings offer an opportunity to study the entire spectrum of processes proceeding in near-and far-field of an SNF repository. The structural geology, mineralogy and petrography, and petrophysical and tectonophysical features of the deposit at its three lower levels are considered. The sequence of metasomatic alteration of rocks and the dynamics of formation of ore-bearing faults that crosscut prototectonic elements, as well as relationships of physicomechanical properties of rocks as a function of the intensity of their metasomatic alteration and the distance from master fault planes, have been established. A 3D geological model of the deposit in combination with estimated parameters of the present-day stress field and physicomechanical properties of particular rock blocks serves as the basis for prediction of the geomechanical behavior of the massif. The practical implications of the results obtained for assessment of the long-term safety of SNF repositories in granites are discussed.

Constraints in the colonization of natural and engineered subterranean igneous rock aquifers by aerobic methane‐oxidizing bacteria inferred by culture analysis

The aerobic methane-oxidizing bacteria (MOB) are suggested to be important for the removal of oxygen from subterranean aquifers that become oxygenated by natural and engineering processes. This is primarily because MOB are ubiquitous in the environment and in addition reduce oxygen efficiently. The biogeochemical factors that will control the success of the aerobic MOB in these kinds of underground aquifers remain unknown. In this study, viable and cultivable MOB occurring at natural and engineered deep granitic aquifers targeted for the disposal of spent nuclear fuel (SNF) in the Fennoscandian Shield (approximately 3-1000 m) were enumerated. The numbers were correlated with in situ salinity, methane concentrations, conductivity, pH, and depth. A mixed population habiting freshwater aquifers (approximately 3-20 m), a potential source for the inoculation of MOB into the deeper aquifers was tested for tolerance to NaCl, temperature, pH, and an ability to produce cysts and exospores. Extrapolations show that due to changing in situ parameters (salinity, conductivity, and pH), the numbers of MOB in the aquifers dropped quickly with depth. A positive correlation between the most probable numbers of MOB and methane concentrations was observed. Furthermore, the tolerance-based tests of cultured strains indicated that the MOB in the shallow aquifers thrived best in mesophilic and neutrophilic conditions as opposed to the hyperthermophilic and alkaliphilic conditions expected to develop in an engineered subterranean SNF repository. Overall, the survival of the MOB both quantitatively and physiologically in the granitic aquifers was under the strong influence of biogeochemical factors that are strongly depth-dependent.

In situ bacterial colonization of compacted bentonite under deep geological high-level radioactive waste repository conditions

Subsurface microorganisms are expected to invade, colonize, and influence the safety performance of deep geological spent nuclear fuel (SNF) repositories. An understanding of the interactions of subsurface dwelling microbial communities with the storage is thus essential. For this to be achieved, experiments must be conducted under in situ conditions. We investigated the presence of groundwater microorganisms in repository bentonite saturated with groundwater recovered from tests conducted at the Aspö underground Hard Rock Laboratory in Sweden. A 16S ribosomal RNA and dissimilatory bisulfite reductase gene distribution between the bentonite and groundwater samples suggested that the sulfate-reducing bacteria widespread in the aquifers were not common in the clay. Aerophilic bacteria could be cultured from samples run at <or=55 degrees C but not at >or=67 degrees C. Generally, the largely gram-negative groundwater microorganisms were poorly represented in the bentonite while the gram-positive bacteria commonly found in the clay predominated. Thus, bentonite compacted to a density of approximately 2 g cm(-3) together with elevated temperatures might discourage the mass introduction of the predominantly mesophilic granitic aquifer bacteria into future SNF repositories in the long run.

Criticality Safety and Sensitivity Analyses of PWR Spent Nuclear Fuel Repository Facilities

Monte Carlo criticality safety and sensitivity calculations of pressurized water reactor (PWR) spent nuclear fuel repository facilities for the Slovenian nuclear power plant Krsko are presented. The MCNP4C code was deployed to model and assess the neutron multiplication parameters of pool-based storage and dry transport containers under various loading patterns and moderating conditions. To comply with standard safety requirements, fresh 4.25% enriched nuclear fuel was assumed. The impact of potential optimum moderation due to water steam or foam formation as well as of different interpretations, of neutron multiplication through varying the system boundary conditions was elaborated. The simulations indicate that in the case of compact (all rack locations filled with fresh fuel) single or “double tiering” loading, the supercriticality can occur under the conditions of enhanced neutron moderation, due to accidentally reduced density of cooling water. Under standard operational conditions the effective multiplication factor (keff) of pool-based storage facility remains below the specified safety limit of 0.95. The nuclear safety requirements are fulfilled even when the fuel elements are arranged at a minimal distance, which can be initiated, for example, by an earthquake. The dry container in its recommended loading scheme with 26 fuel elements represents a safe alternative for the repository of fresh fuel. Even in the case of complete water flooding, the keff remains below the specified safety level of 0.98. The criticality safety limit may however be exceeded with larger amounts of loaded fuel assemblies (i.e., 32). Additional Monte Carlo criticality safety analyses are scheduled to consider the “burnup credit” of PWR spent nuclear fuel, based on the ongoing calculation of typical burnup activities.

Integrated Analytic Radionuclide Transport Model for a Spent Nuclear Fuel Repository in Saturated Fractured Rock

Simple analytic expressions are presented for radionuclide transport from a KBS 3-type repository, where spent nuclear fuel is placed in copper canisters surrounded by bentonite clay and deposited at a depth of 500 m in fractured granitic rock.Dissolution of readily accessible and fuel matrix embedded nuclides, chain decay, and nuclide precipitation is treated within the canister. Transport in the canister void and buffer is modeled with a dual stirred tank analogy, where transport resistances represent an assumed small initial damage in the canister and transport features of the buffer-geosphere interface. Initial, transient diffusion in the buffer is treated with a simple correction term. Chain decay is not included in the buffer.Geosphere transport expressions handle advection, longitudinal dispersion, matrix diffusion, sorption, and radioactive decay, but not chain decay. The treatment is based on earlier results for an instantaneous inlet and for a constant inlet to the geosphere in the nondispersive case. A correction is added so that longitudinal dispersion is taken approximately into account. The correction utilizes analytical expressions for the temporal moments of the geosphere release curve in the dispersive case.The near-field/geosphere integration is treated in a simplified manner avoiding numerical convolutions. The instantaneous inlet expression for the geosphere release is used when the near-field release decreases rapidly in comparison to a typical response time in the geosphere; the constant inlet expression is used in the opposite case.Twenty-seven calculation cases from a safety assessment of a KBS 3 repository using borehole data from three different field investigation sites were repeated with the analytic expressions. The agreement in both near-field and geosphere releases is in general well within an order of magnitude for the variety of long- and short-lived, sorbing, nonsorbing, solubility limited, immediately accessible, and fuel matrix embedded single and ingrowing species dominating the releases and doses in the safety assessment calculations. Also, probabilistic dose calculation results are in good agreement, making the analytic model a versatile complement to the various tools used in long-term safety evaluations of a KBS 3 type of repository in saturated fractured rock.

The SR 97 Safety Assessment of a KBS 3 Repository for Spent Nuclear Fuel – Overview, Review Comments and New Developments

ABSTRACTIn preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Nuclear Fuel and Waste Management Company, SKB has carried out the longterm safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock.The future evolution of the repository system is analysed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion.The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. The results of the assessment also serve as a basis for formulating requirements and preferences regarding the bedrock in site investigations, for designing a programme for site investigations, for formulating functional requirements on the repository's barriers, and for prioritisation of research.SR 97 has been reviewed both by an international group of OECD/NEA experts and by Swedish authorities. The NEA reviewers concluded that “SR 97 provides a sensible illustration of the potential safety of the KBS-3 concept”, and no issues were identified that need to be resolved prior to proceeding to the investigation of potential sites. The authorities' conclusions were in principal consistent with those of the NEA.Uncertainties and lack of knowledge in different areas identified in SR 97 have strongly influenced the contents and structure of SKBs most recent research programme, RD&amp;DProgramme 2001.Since SR 97, the methodology for probabilistic consequence analyses have been further developed. Analytic approximations to the numerical transport models used in SR 97 have been developed. The new models have been used to extend the probabilistic calculations in SR 97.

SR 97: Post-Closure Safety for a KBS 3 Deep Repository for Spent Nuclear Fuel - Overview -

ABSTRACTIn preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Nuclear Fuel and Waste Management Company, SKB has carried out the long- term safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock.The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion.The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. The results of the assessment also serve as a basis for formulating requirements and preferences regarding the bedrock in site investigations, for designing a program for site investigations, for formulating functional requirements on the repository's barriers, and for prioritization of research.