Post-closure Safety Assessment Research Articles

Study on advection–dispersion behavior for simulation of 3H, 99Tc, and 90Sr transport in crushed sandstone of column experiments

Abstract In order to establish a universal and safety-compliant post-closure safety assessment technique, it is necessary to develop appropriate models to explain the migration behavior of radioactive materials within the rock system. Advection-dispersion experiments (ADE) have proven successful in designing transport models through a calibration/validation process of breakthrough curves (BTCs). In the present investigation, we employed a dynamic column device to examine the transport of Tritium (3H), Technetium-99 (99Tc), and Strontium-90 (90Sr) in crushed sandstone. Non-reactive transport experiments utilizing conservative tracers enabled us to determine the transport parameters, including retardation factors (R) and dispersivity (α). Our study focused on investigating the surface complexation model (SCM) using an additive approach coupled with the advection-dispersion equation to simulate the reactive transport behavior of 90Sr. The results affirm the robustness of our chosen thermodynamic database and modeling approach, emphasizing the criticality of accurately modeling and predicting reactive transport behavior in porous media.

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

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

An Assessment of Deep Borehole Disposal Post-Closure Safety

Around the world, deep borehole disposal is being evaluated for intermediate level-waste (ILW), high-level waste, and spent nuclear fuel. To facilitate a disposal concept options analysis for ILW in Australia, desktop and lab-based geoscientific investigations, together with generic post-closure safety assessments of deep borehole disposal of long-lived ILW, have been undertaken. This paper reports on geoscientific data obtained on crystalline rock and rock salt as model rocks for geological disposal. Petrophysical and mineralogical properties for these rocks have been investigated to provide realistic data for evaluation and input to post-closure safety assessments. For crystalline rock samples originating from depths between 700 to 1900 m, very low hydraulic conductivity (2 × 10−12 to 3 × 10−11 m/s) and very low porosity (0.02% to 1.2%) were obtained. The noble gas isotopic composition of fluid inclusions from the same depth interval confirmed the rock had been devoid of recent interaction with meteoric water, thus providing potentially suitable conditions for geological disposal. Rock salt from a 802-m (heterogeneous sample with 40% halite) and a 1100-m (sample with 98% halite) depth also had a low hydraulic conductivity (5 × 10−10 to 5 × 10−9 m/s at 802 m and 10−11 to 2 × 10−10 m/s at 1100 m) and very low porosity (~0.8% for the heterogenous sample and ~0.2% for the pure halite sample). Post-closure safety assessments based on numerical modeling provided bounding conditions around the thermal evolution of the disposal environment in crystalline rock for low heat generating ILW (50 W per 180-L vitrified waste canister), including exploring the sensitivity of temperature evolution within the borehole and rock environment to parameters such as heat load, borehole depth, geothermal gradients, and rock thermal conductivity. The coupling of heat transport with radionuclide migration to account for buoyancy-driven transport was shown to have a limited impact on radionuclide migration. For a disposal borehole in crystalline rock, the radionuclide concentrations and annual dose rates from key radionuclides (99Tc and 79Se) for a 500-m, 1000-m, or 3000-m deep borehole were negligible (i.e., many orders of magnitude smaller than the threshold dose the International Atomic Energy Agency considers insignificant for humans, 0.01 mSv/year). For disposal in rock salt, a suite of numerical model scenarios explored the effectiveness of the engineered barriers, including the glass matrix, primary package, and overpack, assuming diffusion-dominated transport. These scenarios illustrated that the performance of the disposal system was insensitive to the presence or absence of engineered barriers, as dose rates at late time (>105 years) were nearly identical for all scenarios. These results indicate that the natural barrier provided by the salt is very effective at containing radionuclides, while the engineered barriers serve mainly to delay the arrival of the peak dose. While the results are preliminary, the post-closure safety assessments, supported by measured data from crystalline rock and rock salt, give confidence that deep borehole disposal of long-lived ILW would result in dose rates considered insignificant for humans within a few meters from the borehole.

Experimental investigation of uranium release from spent nuclear fuels under conditions expected in a deep geological repository

Abstract. Direct disposal of spent nuclear fuel (SNF) in a deep geological repository is planned for the long-term management of the highly radioactive waste discharged from the light water reactors (LWRs) of nuclear power plants in Germany. The aim is to safely isolate the radionuclide inventory from the biosphere, ad infinitum, making use of a multi-barrier system consisting of technical, geotechnical and geological barriers interposed between the emplaced waste and the environment. Still, reversibility measures including retrievability of the waste products are considered in this concept. Internationally, in safety assessments for deep underground repositories (DGRs), the access of groundwater, followed by failure of canisters and finally loss of the SNF's cladding integrity is considered in the long term. Hence, the evaluation of the performance of SNF in the near field of a deep geological repository requires process understanding of SNF dissolution and rates, in addition to the quantification of the radionuclide release from SNF under reducing conditions of a breached container. With the aim to derive a radionuclide source term, the SNF dissolution and alteration processes can be assigned to two steps, namely the (i) instantaneous release of radionuclides upon cladding failure from gap and grain boundaries and (ii) a long-term release that results from the dissolution of the fuel grains themselves (Ewing, 2015). Since most radionuclides produced during irradiation of nuclear fuel in a reactor are trapped within the SNF matrix, understanding the processes leading to the dissolution of the matrix and its dissolution rates are of particular interest for post-closure safety assessments (Pastina and LaVerne, 2021). Research at the Institute for Nuclear Waste Disposal (INE) at the Karlsruhe Institute of Technology (KIT) has focused, for more than 25 years, on the behaviour of SNF (irradiated UO2 and mixed-oxide, MOX, fuels) under geochemical conditions (pH, redox milieu, gas atmosphere and ionic strength) representative of various repository concepts and radiolytic reactions on SNF dissolution mechanisms. This contribution will give an overview on the radionuclide release from dissolving SNFs under repository conditions. In particular, uranium released into solution from UO2 and MOX fuels during up to 1629 d of leaching in salt brines, granitic/bentonitic or cementitious groundwaters under Ar or Ar/H2 atmospheres. As observed in other published studies on SNF dissolution under reducing conditions, a large scatter of the uranium concentration is seen in the initial stages of our experiments. After about 1 year of leaching, the aqueous concentration of uranium slowly approaches the solubility limit of U(IV), independent of the type of studied SNF samples (various irradiated UO2 and MOX fuels). A similar behaviour is observed for other redox-sensitive actinides. The release rate of uranium decreases significantly in all experiments within first 400 d of leaching. Although the rate is very low, it is >0, and a continuous release is observed. A comparison of the results from SNF experiments under various redox conditions demonstrates that oxidative matrix dissolution is inhibited due to the presence of hydrogen. Our results are discussed in the context of the outcome of international research programmes on SNF corrosion. These findings are considered to be of relevance for the definition of source terms for various irradiated LWR fuels in safety investigations for a DGR in Germany.

Natural Analogue Studies in Support of Post-Closure Safety Assessment of Deep Geological Disposal

Natural analogues are systems that have evolved over geological timescales with features similar to one or several components of a deep geological repository (DGR). Natural analogues complement short-duration laboratory studies since they are existing reflections of many long-term processes that might affect the performance of a repository. Mathematical models are often used for the post-closure safety assessment of a DGR. Confidence in the models’ predictions is enhanced when the models successfully simulate the past evolution of a natural analogue. This paper summarizes the Canadian Nuclear Safety Commission’s (CNSC’s) recent research on natural analogues to inform on (1) glacial erosion, (2) engineered barrier system, and (3) uranium reactive transport in the context of DGRs for radioactive wastes. Glaciation and its erosion are prominent factors impacting the performance of future DGRs at high latitudes in the northern hemisphere. The authors have reviewed the field data from the Greenland Analogue Project, developed a conceptual and mathematical model for the simulation of the thermal conditions within the Greenland ice sheet, as well as the thermal-hydraulic conditions at its base and the ice sheet velocity, and eventually estimated the erosion rate at the site. The Cigar Lake Analogue demonstrates the long-term radionuclide containment capability of the illite clay zone enveloping the ore body, serving as an analogy to the engineered clay barriers. The CNSC and University of Ottawa analyzed 129I in the Cigar Lake core samples, and modeled and correlated the diffusion-dominated transport of radionuclides over the geological evolution of the Cigar Lake deposit. The results provide information on the mobility of fission products and significant radionuclides in conditions analogous to the source, engineered barriers, and near-field host rock of a DGR. The reactive transport and geochemistry of the Kiggavik-Andrew Lake uranium deposit mineralization and remobilization was another natural uranium deposit analogue studied by the CNSC. A reactive transport model was established according to the conceptualized geochemical processes and run under specified boundary and initial conditions to validate the geochemical processes. The geometry, timing, geochemistry, and fluid composition were used as model constraints.

A research and development roadmap to support applications of the enhanced BIOMASS methodology

The International Atomic Energy Agency has coordinated an international project addressing enhancements of methods for modelling the biosphere in post-closure safety assessments of solid radioactive waste disposal. This has resulted in the enhanced BIOMASS methodology that is described elsewhere in this special issue. To a large degree, the enhancements to the BIOMASS methodology arose from experience gained in applying the original methodology, both in the context of other international projects and in assessments of existing or proposed disposal facilities for solid radioactive wastes. Here, this experience is used, together with information on the status of solid radioactive waste disposal programmes worldwide, to identify opportunities for applying the enhanced methodology and for learning from those applications. This provides a basis for identifying research and development to support application of the enhanced methodology in a variety of environmental settings. These research and development requirements include aspects related to climate change under a variety of forcing scenarios, landform development in climatic regimes ranging from cold, polar to arid, tropical, modelling of groundwater flow and contaminant transport in surface-water catchments where both fractured rock and porous sediments are present, and studies of the transport of key radioisotopes of elements central to major biogeochemical cycles, such as those of carbon, chlorine, sulphur and iodine. In addition, some remarks are made on aspects of the application of the enhanced methodology that could imply review and updating of regulations and regulatory guidance, e.g. in relation to the definition of representative persons or groups to be considered in assessments and in respect of approaches to the assessment of radiological impacts on non-human biota. Furthermore, consideration is given as to how the scientific and technical experience that has been gained and methods that have been developed in the context of solid radioactive disposal facilities could support management of contaminated sites and legacy facilities that are likely to require long-term management and stewardship.

Safety assessments undertaken using the BIOMASS methodology: lessons learnt and methodological enhancements

The International Atomic Energy Agency has coordinated an international project addressing enhancements of methods for modelling in post-closure safety assessments of solid radioactive waste disposal. The project used earlier published work from the IAEA biosphere modelling and assessment (BIOMASS) project to further develop methods and techniques. The task was supported by a parallel on-going project within the BIOPROTA forum. The output from the project is described in detail in a forthcoming IAEA report. Here an overview of the work is given to provide researchers in the broader fields of radioecology and radioactive waste disposal with a summarised review of the enhanced BIOMASS methodology and the work that has been undertaken during the project. It is hoped that such dissemination will support and promote integrated understanding and coherent treatment of the biosphere component within the overall assessment process. The key activities undertaken in the project were: review and identification of those parts of the original BIOMASS methodology that needed enhancement, discussions on lessons learned from applying the BIOMASS method, using real examples to assess the methodology and its usefulness, and writing of those parts of the methodology that were considered could benefit from refinement or for which new guidance was required to take account of scientific developments. The work has shown that the overall approach in the original BIOMASS methodology has proven sound. However, the enhanced version clarifies the need for an iterative and holistic approach with system understanding central to the approach. Specifically, experience, especially in site-specific contexts, has emphasised that adequate system understanding is essential in underpinning safety assessments for radioactive waste disposal. The integral role of the biosphere within safety assessment is also emphasised in the enhanced methodology.

Emplacement optimization for the first-stage disposal facility of LILW complex disposal facility by employing simulated annealing algorithm

In order to optimize the emplacement of LILW (Low- and Intermediate-Level radioactive Waste) complex disposal facility, an emplacement optimization program POWER (Program for Optimization of Waste package Emplacement in Repository) based on Simulated Annealing (SA) algorithm is developed. The POWER program derives near-optimum emplacement schemes that minimize the objective function consisting of total dose, radiotoxicity flux, and radiotoxicity concentration obtained from GoldSim calculations for the post-closure safety assessment for LILW rock cavern disposal facility. An optimization problem of sixty nine drums with different degradation mechanism, source term, and drum type to be disposed of into the first-stage disposal facility is defined first and a series of optimization calculations is performed. It is verified that the POWER can generate the near-global optimized emplacement scheme from SA process.

Molecular dynamics study of uranyl adsorption from aqueous solution to smectite

Adsorption of aqueous uranyl species in pore spaces of clay minerals plays a key role in post-closure safety assessments for geological disposal of radioactive waste. Molecular Dynamics Simulations (MDS) were performed to study the adsorption of uranyl (UO22+) to the clay mineral smectite from its aqueous solution in the presence of carbonate (CO32−) and Na+ counter ions. The modelled system consisted of water-saturated clay layers and a 0.162 M aqueous uranyl carbonate solution. The large system size (up to 25,560 atoms) and long simulation times (over 200 ns) allowed investigating the effect of the electrical double layer on uranyl adsorption to smectite for different pore sizes (8.37, 25.1, and 33.5 nm). This study identified various polynuclear uranyl carbonate complexes on clay surfaces (e.g., [Na(UO2)3(CO3)3]+) and in aqueous solution (e.g., [Na2(UO2)5(CO3)6]0) which were not seen in other MDS studies due to statistical limitations caused by the small number of uranyl and carbonate ions in their super cell and/or insufficient simulation time. Uranyl complexes represented the majority of adsorbed species relative to uncomplexed uranyl, with different complexes for different pore sizes. The sorption parameter KD for uranyl ranged from 59 to 151 mL/g and was dependent on pore size (smaller KD for larger pore size). Electrostatic factors controlled the formation and locations of uranyl complexes. MDS-based KD values were in agreement with experimentally derived values for similar experimental conditions. MDS provides an efficient tool to derive sorption parameters for safety assessments, especially in the early stages of site selection and characterisation when access to cores from deep rocks may be limited.

Deep borehole disposal of intermediate-level waste

Abstract. Around the world, deep borehole disposal is being evaluated for intermediate-level waste (ILW), high-level waste (HLW), spent nuclear fuel (SNF), separated plutonium waste and some very high specific activity fission product waste. In Australia, long-lived ILW from research reactors and radiopharmaceutical production represents the principal waste stream that requires deep geologic disposal. Whilst the Australian government has not yet made a decision on its preferred strategy for ILW disposal, deep borehole disposal of small volumes of ILW would be a more cost-effective and modular solution compared to a conventional geologic disposal facility (GDF). CSIRO, ANSTO and SANDIA have created an international partnership to execute a full-scale borehole research, development and demonstration (RD&D) project in Australia. The project will demonstrate the technical feasibility of the long-term safety of borehole disposal in deep geological formations. The execution of this project could also demonstrate options for nuclear waste disposal that would reduce proliferation risks, potentially up to the termination of compliance with international safeguards requirements. The RD&D includes demonstration of surface handling and waste/seal emplacement capabilities, basic research on foundational science areas, and full-scale field testing in both a deep characterization borehole and a larger-diameter (0.7 m or 27.5 inch) 2000 m deep demonstration borehole. The multi-barrier system designed for such a deep disposal borehole concept places much less reliance on engineered barriers at the disposal zone to achieve safety as compared to a conventional GDF. It rather relies on geological features for waste containment. The concept being explored uses disposal containers with primary waste packages, such as vitrified waste canisters, inside; to be both cost effective and fit for purpose, such a container could have a mild steel-based structural component with copper coating. A critical review of six coating technologies showed that cold spray has the greatest advantages, such as minimal porosity and compressive residual stress. The RD&D has delivered novel enabling tools that assist with site screening, borehole design and post-closure safety assessments. For instance, an automated geological fault mapping and meshing tool was developed that assists with ranking the suitability of potential disposal sites based on proximity to faults. New codes were developed for better representation of fault zones in 2D/3D numerical flow and transport models, while also being more efficient to execute. Post-closure safety assessments tested the sensitivity of long-term safety with respect to disposal depth, rock permeability and sorption. Heat transport calculations explored the sensitivity of temperature evolution within the borehole to parameters such as heat load, borehole depth, geothermal gradients and rock thermal conductivity. For verification of host rock tightness while also demonstrating the absence of recent groundwater, a new noble gas analytical facility has been established for measuring rare noble gases in mineral fluid inclusions as indicators of very old pore fluids.

Overview of Post-Closure Criticality Safety – RD&D Topics in Switzerland

Abstract. The demonstration of post-closure criticality safety of spent nuclear fuel in a deep geological repository is a regulatory requirement in Switzerland and many other countries. One of the main challenges stems from the very long timescale (1 million years in Switzerland) that has to be considered. Nagra, the Swiss National Cooperative for the Disposal of Radioactive Waste, is presently elaborating the technical and scientific foundation of the criticality safety assessment in view of the upcoming general licence application for the Swiss Spent Fuel and HLW repository. In this context, Nagra supports and pursues a focussed RD&D programme in collaboration with several renowned research institutes. Nagra's safety concept relies on natural and technical barriers. For the initial thermal phase of the repository, a steel canister assures complete containment of the spent fuel. The canisters are foreseen to remain intact for approximately 10 000 years; however, the subcriticality of the system has to be ensured for a much longer period. In this context, an important part of the research activities pursued by Nagra address the nearfield evolution and the formulation of scenarios for the corresponding evolution of the canister and spent fuel system. The role that variations in the canister design and material composition have on the system's reactivity are also investigated. Other research topics focus on developing a reliable methodology for carrying out the criticality safety assessment. This symposium contribution gives an overview of the post-closure criticality RD&D activities pursued and envisioned by Nagra. The general context and Nagra's fundamental approach to elaborating the current phase of the criticality safety assessment are presented first. Following this, the current RD&D landscape and the most important technical considerations underpinning Nagra's technical basis for the post-closure criticality safety assessment in particular are discussed. Future planned research topics and points of interest are also presented as an outlook of this presentation.

Science underpinning the safety case of deep geological repositories − challenges in the past and in the future and how to maintain knowledge and competence during operation

Final repositories for spent nuclear fuel are approaching implementation. A prerequisite for these advancements is that it has been shown that the repository can be constructed and operated in practice in such a way that safety can be assured both during operation and over very long time scales. The success rests on decades of structured and objective-driven research and development. A key element of the research strategy has been to ensure adequate in-house competence and expertise. Also, openness and international cooperation are essential. Workable procedures for data qualification, version control as well as internal and external peer review have gained importance. When the programmes now enter a new phase of construction and operation new challenges will arise. Even if the implementing organisations would need to keep a core competence on post closure safety assessment international cooperation will be even more important on developing, sharing and managing the knowledge needed.

Climate change and landscape development in post-closure safety assessment of solid radioactive waste disposal: Results of an initiative of the IAEA

The International Atomic Energy Agency has coordinated an international project addressing climate change and landscape development in post-closure safety assessments of solid radioactive waste disposal. The work has been supported by results of parallel on-going research that has been published in a variety of reports and peer reviewed journal articles. The project is due to be described in detail in a forthcoming IAEA report. Noting the multi-disciplinary nature of post-closure safety assessments, here, an overview of the work is given to provide researchers in the broader fields of radioecology and radiological safety assessment with a review of the work that has been undertaken. It is hoped that such dissemination will support and promote integrated understanding and coherent treatment of climate change and landscape development within an overall assessment process.The key activities undertaken in the project were: identification of the key processes that drive environmental change (mainly those associated with climate and climate change), and description of how a relevant future may develop on a global scale; development of a methodology for characterising environmental change that is valid on a global scale, showing how modelled global changes in climate can be downscaled to provide information that may be needed for characterising environmental change in site-specific assessments, and illustrating different aspects of the methodology in a number of case studies that show the evolution of site characteristics and the implications for the dose assessment models.Overall, the study has shown that quantitative climate and landscape modelling has now developed to the stage that it can be used to define an envelope of climate and landscape change scenarios at specific sites and under specific greenhouse-gas emissions assumptions that is suitable for use in quantitative post-closure performance assessments. These scenarios are not predictions of the future, but are projections based on a well-established understanding of the important processes involved and their impacts on different types of landscape. Such projections support the understanding of, and selection of, plausible ranges of scenarios for use in post-closure safety assessments.

Preliminary post-closure safety assessment for a borehole-type repository for disused sealed radioactive sources in Brazil

Brazil has a relatively large inventory of disused sealed radioactive sources (DSRSs). Until now, no decision has yet been made about the final destination of this category of radioactive wastes, although a repatriation of a small fraction of these sources comprising mainly neutron and high activity sources was already carried out. Borehole type repositories are one disposal solution considered for DSRSs in Brazil. This paper addresses a preliminary post-closure safety assessment for such a facility, using the borehole disposal concept (BDC) applied to different geological conditions and a range of projected inventories. Results from running the AMBER code considering deterministic and stochastic approaches showed that Am-241 is the main source of potential concern in order to comply with the effective dose constraint of 0.3 mSv/y and allowed the establishment of the relation between the maximum Am-241 inventory and the hydraulic conductivity of the geosphere.

Characterisation of a high pH cement backfill for the geological disposal of nuclear waste: The Nirex Reference Vault Backfill

In a conceptual UK geological disposal facility for nuclear waste within a high-strength, crystalline geology, a cement-based backfill material, known as Nirex Reference Vault Backfill (NRVB), will be used to provide a chemical barrier to radionuclide release. The NRVB is required to have specific properties to fulfil the operational requirements of the geological disposal facility (GDF); these are dependent on the chemical and physical properties of the cement constituent materials and also on the water content. With the passage of time, the raw materials eventually used to synthesise the backfill may not be the same as those used to formulate it. As such, there is a requirement to understand how NRVB performance may be affected by a change in raw material supply. In this paper, we present a review of the current knowledge of NRVB and results from a detailed characterisation of this material, comparing the differences in performance of the final product when different raw materials are used. Results showed that minor differences in the particle size, surface area and chemical composition of the raw material had an effect on the workability, compressive strength, the rate of hydration and the porosity, which may influence some of the design functions of NRVB. This study outlines the requirement to fully characterise cement backfill raw materials prior to use in a geological disposal facility and supports ongoing assessment of long-term post-closure safety.

폐밀봉선원 처분시스템 예비 폐쇄후 안전성평가

An optimum disposal plan of disused sealed radioactive sources (DSRSs) should be established to ensure long-term disposal safety at the low- and intermediate-level radioactive waste (LILW) disposal facility in Gyeongju. In this study, an optimum disposal system was suggested and preliminary post-closure safety assessment was performed. The DSRSs disposal system was composed of a rock cavern and near surface disposal facilities at the Gyeongju LILW disposal facility. The assessment was conducted using GoldSim program, and probabilistic assessment and sensitivity analysis were implemented to evaluate the uncertainties in the input parameters of natural barriers. Deterministic and probabilistic calculations indicated that the maximum dose was below the regulatory limits (<TEX>$0.1mSvyr^{-1}$</TEX> for the normal scenario, <TEX>$1mSvyr^{-1}$</TEX> for the well scenario). It was concluded that the DSRSs disposal system would maintain environmental safety over a long-time. Moreover, the partition coefficient of Np in host rock, Darcy velocity in host rock, and density of the host rock were the most sensitive parameters in predicting exposure dose in the safety assessment.

The Swedish radiological environmental protection regulations applied in a review of a license application for a geological repository for spent nuclear fuel

For the first time, a system for specific consideration of radiological environmental protection has been applied in a major license application in Sweden. In 2011 the Swedish Nuclear Fuel & Waste Management Co. (SKB) submitted a license application for construction of a geological repository for spent nuclear fuel at the Forsmark site. The license application is supported by a post-closure safety assessment, which in accordance with regulatory requirements includes an assessment of environmental consequences. SKB's environmental risk assessment uses the freely available ERICA Tool. Environmental media activity concentrations needed as input to the tool are calculated by means of complex biosphere modelling based on site-specific information gathered from site investigations, as well as from supporting modelling studies and projections of future biosphere conditions in response to climate change and land rise due to glacial rebound. SKB's application is currently being reviewed by the Swedish Radiation Safety Authority (SSM). In addition to a traditional document review with an aim to determine whether SKB's models are relevant, correctly implemented and adequately parametrized, SSM has performed independent modelling in order to gain confidence in the robustness of SKB's assessment. Thus, SSM has used alternative stylized reference biosphere models to calculate environmental activity concentrations for use in subsequent exposure calculations. Secondly, an alternative dose model (RESRAD-BIOTA) is used to calculate doses to biota that are compared with SKB's calculations with the ERICA tool. SSM's experience from this review is that existing tools for environmental dose assessment are possible to use in order to show compliance with Swedish legislation. However, care is needed when site representative species are assessed with the aim to contrast them to generic reference organism. The alternative modelling of environmental concentrations resulted in much lower concentrations compared to SKB's results. However, SSM judges that SKB's in this part conservative approach is relevant for a screening assessment. SSM also concludes that there are big differences in dose rates calculated to different organisms depending on which tool that is used, although not systematically higher for either of them. Finally, independent regulatory modelling has proven valuable for SSM's review in gaining understanding and confidence in SKB's assessment presented in the license application.

A method for independent modelling in support of regulatory review of dose assessments

Several countries consider geological disposal facilities as the preferred option for spent nuclear fuel due to their potential to provide isolation from the surface environment on very long timescales. In 2011 the Swedish Nuclear Fuel & Waste Management Co. (SKB) submitted a license application for construction of a spent nuclear fuel repository. The disposal method involves disposing spent fuel in copper canisters with a cast iron insert at about 500 m depth in crystalline basement rock, and each canister is surrounded by a buffer of swelling bentonite clay. SKB's license application is supported by a post-closure safety assessment, SR-Site. SR-Site has been reviewed by the Swedish Radiation Safety Authority (SSM) for five years. The main method for review of SKB's license application is document review, which is carried out by SSM's staff and supported by SSM's external experts. The review has proven a challenging task due to its broad scope, complexity and multidisciplinary nature. SSM and its predecessors have, for several decades, been developing independent models to support regulatory reviews of post-closure safety assessments for geological repositories. For the review of SR-Site, SSM has developed a modelling approach with a structured application of independent modelling activities, including replication modelling, use of alternative conceptual models and bounding calculations, to complement the traditional document review. This paper describes this scheme and its application to biosphere and dose assessment modelling. SSM's independent modelling has provided important insights regarding quality and reasonableness of SKB's rather complex biosphere modelling and has helped quantifying conservatisms and highlighting conceptual uncertainty.

Post-closure safety assessment of near surface disposal facilities for disused sealed radioactive sources

Great attention has been recently paid to the post-closure safety assessment of low- and intermediate-level radioactive waste (LILW) disposal facility for disused sealed radioactive sources (DSRSs) around the world. Although the amount of volume of DSRSs generated from industry, medicine and research and education organization was relatively small compared with radioactive wastes from commercial nuclear power plants, some DSRSs can pose a significant hazard to human health due to their high activities and long half-lives, if not appropriately managed and disposed. In this study, post-closure safety assessment was carried out for DSRSs generated from 1991 to 2014 in Korea in order to ensure long-term safety of near surface disposal facilities. Two kinds of disposal options were considered, i.e., engineered vault type disposal facility and rock-cavern type disposal facility. Rock-cavern type disposal facility has been under operation in Gyeongju city, republic of Korea since August 2015 and engineered vault type disposal facility will be constructed until December 2020 in the vicinity of rock-cavern disposal facility. Assessment endpoint was individual dose to the member of critical group, which was modeled by GoldSim, which has been widely used as probabilistic risk analysis software based on Monte Carlo simulation in the area of safety assessment of radioactive waste facilities. In normal groundwater scenario, the maximum exposure dose was extremely low, approximately 1×10−7mSv/yr, for both disposal options and satisfied the regulatory limit of 0.1mSv/yr. However, in the drinking well scenario, the maximum exposure dose for engineered vault type disposal facility was assessed as 2.022mSv/yr where the value exceeded the regulatory limit of 1mSv/yr. The maximum exposure dose for rock-cavern type disposal facility was calculated to be 0.634mSv/yr, whose results was relatively very close to the regulatory limit considering high uncertainty of long-term environmental conditions. It was demonstrated that DSRSs including the radionuclides of 14C, 226Ra and 241Am significantly contributed to the large portion of exposure dose to the public based on the long-term safety assessment. Therefore, it was recommended that the near surface disposal of DSRSs containing 14C, 226Ra and 241Am should be restricted and managed by long-term interim storage option in order to minimize their potential radiological health effects.

Long-term safety of the extended SFR - Methodology and conclusions from the SR-PSU project

ABSTRACTSince 1988 the Swedish Nuclear Fuel and Waste Management Co. operates a repository for low- and intermediate-level short-lived radioactive waste, SFR, in Forsmark, Sweden. Due to decommissioning of the nuclear power plants additional storage capacity is needed. In December 2014, an application to extend the repository was therefore submitted. One key component of this application was an assessment of post-closure safety of the extended SFR. For this safety assessment, a methodology based on that developed by SKB for the spent nuclear fuel repository was used and the impact of the degradation of repository components, the evolution of the surface system and changes of future climate on the radiological safety of the repository was assessed over a period of 100,000 years. The central conclusion of the SR-PSU safety assessment is that the extended SFR repository meets requirements on protection of human health and of the environment that have been established by the Swedish radiation safety authority for the final disposal of radioactive waste. Furthermore, the design of the repository was shown suitable for the waste selected and the applied methodology suitable for the safety assessment.