Disposal Of Spent Nuclear Fuel Research Articles

Technical, Regulatory, and Policy Feasibility Assessments of Rod Consolidation in Spent Nuclear Fuel Disposal

Various approaches have been proposed to address the challenges associated with storage and disposal of spent nuclear fuel. Rod consolidation, which involves extracting only the fuel rods from fuel assemblies and combining them for storage with reduced spacing, gained attention until the 1990s for its benefits in storage efficiency. In this case study, experts from diverse backgrounds evaluate the technical and political feasibility of rod consolidation for enhancing disposal practices. This study conducts a technical and regulatory analysis of rod consolidation under disposal conditions and evaluates its feasibility from a disposal perspective. From a technical viewpoint of 2:1 and 1.5:1 rod consolidations, the rod consolidation methods demonstrate positive effects on nuclear criticality compared to the reference, non-rod consolidation case. For the advanced repository design with elevated temperature conditions of 150 °C or higher, the applicability of rod consolidation becomes more convincing, thereby achieving a reduction in the number of required disposal canisters. In addition, regulatory considerations primarily focus on the requirements for applying rod consolidation technology under the Korea Nuclear Act. Moreover, this research conducts environmental, economic, and acceptance analyses to ascertain the viability of rod consolidation. The findings provide compelling evidence supporting the feasibility of rod consolidation technology for disposal purposes, highlighting its technical, regulatory, economic, environmental, and societal acceptance advantages. However, challenges remain, particularly concerning its applicability to high-burnup nuclear fuel and unresolved regulatory and technical issues.

Cross-jurisdictional analysis and forecasting of North American nuclear fuel inventory using a standardized unit

This study fills the noted gap in comparative analyses of spent nuclear fuel (SNF) by assessing inventories from two key nuclear power regions, the USA and Canada, using a comprehensive analytical framework and standardized data from 2009 to 2021. In the USA, SNF inventory increased by 14.7 % in fuel assembly weight and 47 % in residual heavy metal content compared to Canada, in line with their use of light water reactors. Canada's SNF production is directly correlated to its nuclear power output, influenced by the lower burnup of natural uranium fuel used in CANDU reactors (R2 = 0.57; p-value < 0.05) while the USA shows insignificant correlation, likely due to a variety of reactor types and higher burnup rates (R2 = 0.008; p-value > 0.05). Further, the study identifies a strong negative correlation between uranium mine production and SNF inventory in the USA, indicating a reliance on imports amidst negligible domestic mining. In contrast, Canada also exhibits moderate negative dependency due to its position as a major uranium exporting jurisdiction. The obtained negative correlations with coal rents in both countries indicate a shift towards more nuclear energy use, impacting economic growth and energy patterns. The developed predictive models indicate a higher future SNF increase in Canada than in the USA. These findings are essential for planning the transition from temporary to permanent SNF disposal, ensuring safe long term management of radioactive waste.

Coupled 3D thermal-hydraulic code development for performance assessment of spent nuclear fuel disposal system

As a solution to the problem of spent nuclear fuels (SNFs), the disposal of SNF has gained attention from nations using nuclear energy because of hazards posed to the ecosystem. Among many proposed solutions, the most promising method is to dispose of SNF in a deep geological repository (DGR) which utilizes the multi-barrier concept developed by Finland and Sweden. Here, a new fully-coupled Thermal-Hydraulic (TH) code HADES (High-level rAdionuclide Disposal Evaluation Simulator) is developed using the MOOSE framework. This new code suggests basic numerical tools, such as a non-linear solver and finite element discretization, to assess the safety performance of disposal systems. The new TH code considered various TH behavior using Richards' flow approach, assuming gas pressure is constant. The HADES showed promising results when it was compared to various TH codes validated from DECOVAELX-THMC projects. When the single-canister model was utilized to estimate the TH behavior of the Korean Reference disposal System, although it showed significant saturation reduction due to the evaporation of water, the temperature was maintained under the thermal criteria limit, which is 100 °C. In addition, the new code estimated temperature and degree of saturation of the multi-canisters model, considering two or three canisters, it showed a slightly lower temperature, 5 °C, than the single-canister model. From these results, the following are concluded: (1) the new TH code contribute to an additional integrity by estimating TH behavior of KRS; (2) however, due to limitations in single-canister simulation, it is recommended to use multi-canisters simulation to estimate TH behavior accurately. Therefore, this model is anticipated not only to help licensing applications and estimation of various multi-physics phenomena and multi-canister at the disposal site.

Preliminary Selection of Safety-Relevant Radionuclides for Long-Term Safety Assessment of Deep Geological Disposal of Spent Nuclear Fuel in South Korea

Preliminary Selection of Safety-Relevant Radionuclides for Long-Term Safety Assessment of Deep Geological Disposal of Spent Nuclear Fuel in South Korea

Permeability and diffusion of tritium in bentonite

Abstract. As a part of the Croatian RD&amp;amp;D (research, development, and demonstration) program within the Program for LILW (low- and intermediate-level waste) and SNF (spent nuclear fuel) disposal, research of permeability and the diffusion of tritium in bentonite was organized, as part of the research of the future engineering barriers for repositories of LILW and SNF. Two types of materials were prepared, namely pure sodium bentonite and a mixture of sodium bentonite (12 %) with quartz sand (88 %). Compaction of materials was conducted with a modified Proctor test, and the apparatus was designed and made at the Faculty of Mining, Geology and Petroleum Engineering at the University of Zagreb, utilizing fused deposition modeling 3D printer and PETG (polyethylene terephthalate glycol) material. Water enriched with 3H has been prepared by dilution of the National Institute of Standards and Technology (NIST) 3H standard (4926e) to 6 L. The activity of the prepared enriched water was 49 700±180 TU. One part of the device was filled with 3H-free water and the other with 3H-spiked water (0.6 L volume for each part). In total, five devices with bentonite and five devices with a sand–bentonite mixture were prepared this way. The devices were thoroughly sealed to avoid loss of sample and isotopic exchange with the atmosphere and put horizontally in holders. The water for 3H activity measurements was sampled monthly or bimonthly. A total of 8 mL of the sample was taken by syringe and transferred to a measurement vial. Each vial was additionally filled with 12 mL of a scintillator (Ultima Gold LLT, PerkinElmer) and thoroughly mixed. 3H activity was determined by a liquid scintillation counter (LSC; Quantulus 1220) at the Laboratory for Low-level Radioactivities, Ruđer Bošković Institute, Zagreb. Due to the penetration of bentonite from the barrier into the water samples, tritium-free water has relatively high activity after 2 months of the experiment (average value was 5560±240 TU). In the final sampling, after a 2-month time span, the activity reached 9670±2490 TU. Later, sampling and measurement were not possible due to the high bentonite amount within the samples. Tritium-free samples from the sand–bentonite mixture device were stable, and the first analyses (1 month after the experiment setup) showed an average activity of 533±31 TU, while the final (5-month time span from the first sampling) average activity was 12 800±3430 TU. The increase in the concentration from the beginning to the end of the experiment was almost exponential. Spiked samples from the bentonite device had an average activity of 36 950±2550 TU at the first sampling and 17 580±13 320 TU after a 1-month time leg. Later, sampling was not possible. Spiked samples from sand–bentonite mixture showed a decrease in tritium activity from the first sampling (43 390±1310 TU) to 27 210±285 TU (5 months after the first sampling). Considering the penetration of bentonite material into the water samples, only the sand–bentonite mixture can be considered for further analyses through the calculation of the diffusion.

Review on the Measurement of Thermal Properties of Rocks for Deep Geological Disposal of Spent Nuclear Fuel

Review on the Measurement of Thermal Properties of Rocks for Deep Geological Disposal of Spent Nuclear Fuel

Sensitivity study of risk to future generations from geological disposal of radioactive waste in high strength rock in the UAE

The United Arab Emirates (UAE) Government's Policy is to institute a safe and permanent geological disposal facility (GDF) for the final disposal of higher activity radioactive wastes (HAW). The UAE has yet to identify or decide on a disposal site, and geological data specific to the characterisation of a GDF host rock either lacks sufficient details or is not available in the public domain. However, initial screenings of geological environments in the UAE give an indication of potential suitable host rock formations. This paper presents the results of a study into the factors that influence the risk to future generations from the disposal of Spent Nuclear Fuel (SNF) in a deep underground geological disposal facility. For the purpose of this study the GDF was assumed to be in a High Strength Rock (HSR) geological environment in the Khor Fakkan (KF) Block. GoldSim's Contaminant Transport code was used to model the potential dispersion of radioactivity from a GDF. which enabled the calculation of risk to future generations. The GDF radiological source term was based on the UAE's forecasted SNF inventory. The model created from GoldSim emphasizes the transport of radionuclides through the GDF's engineered barrier systems (EBS), through the surrounding geosphere and into the biosphere. Once in the biosphere the radiation dose to the reference receptor is determined via ingestion from well drinking water, or inhalation through atmospheric dispersion. The paper explores the factors that influence the assessment of risk to people both now and in the future. The analyses presented in this paper are not meant to be used in site selection decision making or in the assessment of the suitability of the location to host a GDF for nuclear waste. The selection of a disposal site in the UAE would depend on a more detailed analysis of site-specific conditions.

Sensitivity study of risk to future generations associated with the Geological Disposal of Radioactive Waste in the Jebel-Al-Dhanna salt dome

The disposal of Higher Activity Wastes, particularly long lived Heat Generating High Level Waste (HG-HLW) such as Spent Nuclear Fuel (SNF) is one of the biggest challenges encountered by the peaceful civil nuclear energy industry. The United Arab Emirates (UAE) has emplaced laws and instituted a policy in which geological disposal is a potential choice for SNF disposal. Available geological data in the public domain provide initial indications of geological environments that might be suitable to host a Geological Disposal Facility (GDF) in the UAE. This paper aims to present an indicative preliminary Performance Assessment (PA) demonstrating radiological dose and risk estimations from hosting a GDF in the Jebel al Dhanna (JAD) location consisting of an Evaporite geological environment in the form of a Salt Dome (Diapir) that protrude above the surface. Modelling the dispersion of harmful radionuclides from the GDF hosted at the JAD location is achieved by utilising GoldSim's Contaminant Transport code, which has successfully provided an evaluation of radiological risks to current and future generations from the disposal of the UAE's forecasted SNF inventory. The developed GoldSim model highlights the migration of radionuclides from within disposed waste packages, past the engineered barrier system (EBS), through the surrounding geosphere or the Natural Barrier System (NBS), reaching a reference receptor. Due to the lack of detailed and site specific data, analyses presented in this paper are to be viewed as generic and are not to be utilised directly as a site selection tool. The analyses presented are expected to go through several revisions and development stages once sufficient site specific data is made available and will then be presented in, and be a part of a UAE site specific safety case report. The analyses presented in this paper are not meant to be used in site selection decision making or in the assessment of the suitability of the location to host a GDF for nuclear waste. The selection of a site for a GDF in the UAE is a matter for the UAE Government and depend upon a more detailed analysis of site-specific conditions.

The Art of Being Ethical and Responsible: Print Media Debate on Final Disposal of Spent Nuclear Fuel in Finland and Sweden

After decades of preparation, the final disposal of spent nuclear fuel has reached the construction stage in Finland, and the neighboring Sweden is likely to soon follow in the footsteps. These Nordic countries rely on a similar technical concept based on passive safety, advocated as a means of minimizing the burden to future generations. The scholarly literature on the ethics of nuclear waste management has thus far paid little attention to the views of the broader publics on the associated ethical challenges. This article helps to fill the gap through a longitudinal and comparative analysis of ethical discussion of the final disposal of SNF in news articles and letters to the editor in four leading Finnish and Swedish daily newspapers in 2008–2015. The study period included major milestones in the licensing processes of the respective two repository projects. The article examines the attention paid to intra- and intergenerational distributive and procedural justice, the changes in the ethical agenda over time, and the societal actor groups that receive attention in the media. The analysis reveals two distinct ethical media agendas: (1) the news article agenda that is dominated by framings of the main players (industry, politicians, authorities, and experts) and largely excludes future generations from the scope of justice, and (2) the agenda represented by the letters to the editor, which focuses on intergenerational justice concerns. Particularly, in the Finnish letters to the editor the value of the lives of distant future generations was discounted implicitly.

Evaluation of the corrosion behavior of modern spent nuclear fuels under repository conditions

Abstract. 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&amp;amp;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).

Slovak legal system for ensuring feasible nuclear back-end system implementation

In the 70-plus years of peaceful uses of nuclear energy, the nuclear industry has reached a mature awareness of its importance, influence and responsibility for and towards the world. Not only does it dutifully regulate itself in technical aspects, making the use of nuclear energy stricter and safer by any and all means possible, it also establishes policies, backed by legislation, for its decades-long activities. Thus, the industry has recognised the need to systematically approach and solve the issue of complete dismantling of its (previously operating) facilities and final disposal of the operational and decommissioning waste (including spent nuclear fuel (SNF)1). This sought end state must be reached both by technical means as well as by creating a sound system for today’s continuous accumulation of funds, needed for the future implementation of the selected technical solution (including final disposal of SNF).

Role of the host communities in final disposal of spent nuclear fuel in Finland and Sweden

Finland and Sweden are the countries with the most advanced plans for final disposal of spent nuclear fuel (SNF). Both countries have also been evoked as good examples in the use of a 'partnering' approach, designed to achieve both a licensable site supported by the community and a balance between fair representation and competent participation. While both are consensus-driven high-trust societies, with similar technological concept for SNF disposal, and whose licensing processes have advanced at a fairly similar pace, their nuclear waste policies also differ from each other in certain key aspects. One difference concerns the role of the communities in siting and licensing of the repositories. The paper examines 1) the background for this differentiation, 2) how local final disposal organizations in the host communities, Eurajoki and Östhammar, took shape and evolved, and 3) how differences between the organizations illustrate the divergence between the Finnish and Swedish approaches to stakeholder engagement. While the Swedish approach can be characterized as 'involved partnership' – which shapes the operating environment for the implementer and authorities by challenging and even modifying policies and actions – the Finnish case could be described as a 'bystander partnership' characterized by trust in safety authorities, with community economics as the primary concern.

Post-Closure Safety Calculations for the Disposal of Spent Nuclear Fuel in a Generic Horizontal Drillhole Repository

The post-closure performance of a generic horizontal drillhole repository for the disposal of spent nuclear fuel (SNF) is quantitatively evaluated using a physics-based numerical model that accounts for coupled thermal-hydrological flow and radionuclide transport processes. The model incorporates most subcomponents of the repository system, from individual waste canisters to the geological far field. The main performance metric is the maximum annual dose to an individual drinking potentially contaminated water taken from a well located above the center of the repository. Safety is evaluated for a wide range of conditions and alternative system evolutions, using deterministic simulations, sensitivity analyses, and a sampling-based uncertainty propagation analysis. These analyses show that the estimated maximum annual dose is low (on the order of 10−4 mSv yr−1, which is 1000 times smaller than a typical dose standard), and that the conclusions drawn from this dose estimate remain valid even if considerable changes are made to key assumptions and property values. The depth of the repository and the attributes of its configuration provide the main safety function of isolation from the accessible environment. Long-term confinement of radionuclides in the waste matrix and slow, diffusion-dominated transport leading to long migration times allow for radioactive decay to occur within the repository system. These preliminary calculations suggest that SNF can be safely disposed in an appropriately sited and carefully constructed and sealed horizontal drillhole repository.

Siting Deep Boreholes for Disposal of Radioactive Waste: Consequences for Tight Coupling between Natural and Engineered Systems.

Since the Yucca Mountain project in the U.S. was defunded in 2010, the notion of disposing of spent nuclear fuel (SNF) in deep boreholes has been reinvigorated, most recently by private companies proposing to utilize lateral drilling technology to excavate boreholes for SNF disposal in sedimentary rock. It is claimed that this approach will alleviate site characterization efforts and expand the availability of potential disposal sites. However, long-term safety will hinge upon the prevalence of geochemically reducing, highly saline, and slow-flowing fluids around the waste emplacement zone, and to quantify these parameters in fluids sampled from depths >1 km will present a challenge. Regional data indicate only a narrow geographical extent of such conditions in the conterminous United States. Furthermore, models of radionuclide transport from disposal boreholes must take into account processes that may accelerate degradation of the canisters, plug, and SNF itself, such as radiolysis and attack by constituents of hydrothermal brines, coupled with hydrogeologic features that promote advective groundwater flow. This review summarizes some geologic considerations, most notably those related to geochemistry, that challenge the long-term safety case for deep borehole disposal of SNF.

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.

Additional Capability Being Developed from UNF-ST&DARDS Unified Database: System Analysis and Fuel Cycle Option Analysis

The Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) is being developed for the U.S. Department of Energy’s Office of Nuclear Energy by the national laboratories. An important part of UNF-ST&DARDS is the Unified Database (UDB), which contains information that can support a variety of activities including fuel storage, fuel transportation, and disposal-related system analysis. Currently, the main application of the UDB is to support evaluation of the characteristics of discharged spent nuclear fuel (SNF) from the U.S. commercial reactors. However, because of the extensive amount of data that has been collected and analyzed for UNF-ST&DARDS, there are many more applications that can utilize the UDB including system analysis with the Next-Generation System Analysis Model (NGSAM) and fuel cycle analysis with fuel cycle simulation codes such as ORION. Going forward, NGSAM and fuel cycle transition analysis with ORION integrate UDB data wherever possible in the UDB’s development plan. These advances in NGSAM and fuel cycle analysis can be used in conjunction with the UDB to help answer more complex questions about the optimization, utilization, storage, and eventual disposal of SNF.

Assessment of spent nuclear fuel amounts to be managed based on disposal option in Republic of Korea

The Republic of Korea (ROK) was operating 24 nuclear power plants (NPPs) at the end of 2015 and nearly 30% of electricity was supplied by the fleet of NPPs. Currently, spent nuclear fuel (SNF) from the NPPs is stored in pools attached to each NPP. However, the pools have been gradually filling up and would be saturated within decades. Recently, The Public Engagement Commission on Spent Nuclear Fuel Management (PECOS) submitted its final report to government proposing direct disposal of SNF from the early 2050s. It also suggested construction of an interim storage facility at the same site. In this work, the amount of SNF was quantified based on various nuclear scenarios. National electricity plans address the installed capacity of nuclear to 2035. To extend the period of the scenarios up to 2100, the results of open research were adopted and nine sub-scenarios were derived, considering the prospect for electricity demand (highest, reference and lowest) as well as the lifetime extension of NPPs. The total amount of SNF was estimated at between 60 and 102ktHM at 2100. The required at-reactor (AR) storage capacity was evaluated within a range of 10–12ktHM and an away-from-reactor (AFR) facility would store 14–17ktHM depending on the scenarios.

Feasibility Studies on Pyro-SFR Closed Fuel Cycle and Direct Disposal of Spent Nuclear Fuel in Line with the Latest National Policy and Strategy of Korea

With a view to providing supportive information for the decision-making on the direction of the future nuclear energy systems in Korea (i.e., direct disposal or recycling of spent nuclear fuel) to be made around 2020, quantitative studies on the spent nuclear fuel (SNF) including transuranic elements (TRUs) and a series of economic analyses were conducted. At first, the total isotopic inventory of TRUs in the SNF to be generated from all thirty-six units of nuclear power plants in operation or under planning is estimated based on the Korean government’s official plan for nuclear power development. Secondly, the optimized deployment strategies are proposed considering the minimum number of sodium cooled-fast reactors (SFRs) needed to transmute all TRUs. Finally, direct disposal and Pyro-SFR closed nuclear energy systems were compared using equilibrium economic model and considering reduction of TRUs and electricity generation as benefits. Probabilistic economic analysis shows that the assumed total generation cost for direct disposal and Pyro-SFR closed nuclear energy systems resides within the range of 13.60~33.94 mills/kWh and 11.40~25.91 mills/kWh, respectively. Dominant cost elements and the range of SFR overnight cost which guarantees the economic feasibility of the Pyro-SFR closed nuclear energy system over the direct disposal option were also identified through sensitivity analysis and break-even cost estimation.

Is there a Nordic model of final disposal of spent nuclear fuel? Governance insights from Finland and Sweden

This paper explores citizen participation in Swedish and Finnish regulatory processes for final disposal of spent nuclear fuel (SNF). Finland and Sweden are considered the most advanced worldwide in term of SNF disposal plans. Our aim is to analyze the institutional waste management frameworks, focusing on the role of civil society organizations (CSOs); how lay-people and civil society organizations have been able to participate and contribute to radioactive waste licensing processes; and the nature of radioactive waste risk debates. We review official documents of the waste companies and nuclear safety authorities, plus information from civil society organizations and laypeople. Our theoretical framework takes a civil regulation perspective, which is oriented towards institutional issues. The analysis indicates that civil regulation of SNF is better established in Sweden than in Finland because of institutional arrangements and the more controversial nature of nuclear power. Swedish civil regulation resembles a more liberal approach, whereas in the Finnish case technocratic domination is more evident and therefore CSOs have been left to choose critical, confrontational and antagonist civil regulation strategies producing dispersed and random civil regulation. We conclude that due to differences in civil regulation there is not a Nordic model for SNF disposal.

The Regulations of Japan on the Facility for the Disposal of Spent Nuclear Fuel and the Implications Thereof

The Regulations of Japan on the Facility for the Disposal of Spent Nuclear Fuel and the Implications Thereof