Spent Nuclear Fuel Management Research Articles

Development of Critical Thinking in Chemistry Teaching

The critical thinking is an important part of education supported by the Czech curriculum documents, and its importance is now being discussed in the context of the development of artificial intelligence. In the teaching of (not only) chemistry, the critical thinking can be purposefully developed through appropriate activities. The article presents three comprehensive teaching tasks for chemistry education that have the potential to develop critical thinking: Genetically Modified Crops (a task using the problem-based learning and analysis of information sources), Nuclear Energy and Spent Nuclear Fuel Management (a task focused on critical reading of text, reasoning and working with artificial intelligence) and Reaction Kinetics (research-oriented laboratory work enhanced with elements of graph and data interpretation and reasoning). Building critical thinking skills in chemistry is essential for understanding abstract concepts and acquiring skills that enable students to approach chemistry topics responsibly and knowledgeably in everyday life. Full text English translation is available in the on-line version.

Enhanced formula-based method to estimate burnup, initial enrichment, and cooling time using limited gamma-ray source data of reference PWR spent nuclear fuel assemblies

This study proposes a framework to develop a model for estimating burnup (BU), initial enrichment (IE), and cooling time (CT) of spent nuclear fuel (SNF) overcoming the limitations of existing models in determining IE. Database (DB) of gamma-ray sources for PWR SNF was generated using ORIGAMI. Target gamma radionuclides were selected based on the ease and credibility of measurement. The concept of Product of Powers of Radioactivity (PPR) was suggested, and PPRs for modeling were selected through evaluation of all possible PPRs. Functions established for PPRs were applied to estimate BU, IE, and CT (BIC) of PWR SNF assemblies. Optimized PPR combinations for estimating BU, IE, and CT respectively, as well as three variables simultaneously with the highest accuracy, were suggested. The estimation model proposed in this study enhanced IE estimations, maintaining the reliability of BU and CT estimations. The above framework enables construction of intuitive BIC estimation models using only a small amount of data. This can help hasten the development of estimation models for SNF from various reactor types. Consequently, it is expected to contribute to the safety and safeguards of SNF management by improving the verification that SNF meets acceptance criteria and conforms to the declared information.

A Framework to Assess Advanced Reactor Spent Fuel Management Facility Deployment

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. The back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

Assessing the Potential of Small Modular Reactors (SMRs) in Spent Nuclear Fuel Management: A Review of the Generation IV Reactor Progress

Assessing the Potential of Small Modular Reactors (SMRs) in Spent Nuclear Fuel Management: A Review of the Generation IV Reactor Progress

Exploring Public Discourse About Spent Nuclear Fuel Management on Twitter

Twitter provides a complementary source of information to more traditional mechanisms of data collection of public opinion about technically complex and controversial topics like spent nuclear fuel (SNF) management. In this study, we apply content and network analysis to investigate the discourse surrounding public concerns about SNF management in the United States on Twitter during 2021. We explore two key research questions: (1) What issues do individuals bring up in the messages about SNF on social media platforms like Twitter? and (2) What can we learn from the network structure of online conversations about SNF management on Twitter? The results indicate that tweets about technology, community, health and safety, and legal and political topics consistently outpaced concerns about economic and environmental impacts connected to SNF management on Twitter, emphasizing potential risks more than benefits. This study also found that Twitter users’ conversations about SNF occur in tightly clustered groups of users centered around a small number of accounts. Together, the content and structure of the online discourse offers insight into public attitudes and concerns about SNF management from a variety of sources, providing experts with a baseline analysis of potentially salient issues that can inform communication with public audiences.

Assay data of spent nuclear fuel: the lab-work behind the numbers

Computational modelling for spent nuclear fuel (SNF) characterization is already widely used and continuously further developed for a plethora of safety related applications and licensing issues in SNF management. An essential step in the development of these methodologies is the validation: the demonstration that the SNF elemental and isotopic composition is sufficiently accurately predicted by the code calculations. This validation step requires experimentally measured nuclide concentrations in SNF, together with an estimation of related uncertainties. The SFCOMPO 2.0 database of OECD/NEA is a database of such publicly available assay data of SNF. A basic understanding of all analytical steps that finally result in assay data of SNF is important for modelers when assessing the “fit-for-validation” requirement of an experimental dataset. The aim of this article is to explain users of such datasets the complex analytical pathway towards assay data. Points of attention, challenges and pitfalls all along the analytical pathway will be discussed, from sampling, dissolution procedures, necessary dilutions and separations, available analytical techniques, some related uncertainties, up to reporting of the results.

Application of the Analytic Hierarchy Process in selecting long-term storage option for spent nuclear fuel in South Africa

South Africa is one of the countries using nuclear power and has prospects to expand its use in the near future. The major issue with nuclear power generation is the management of spent nuclear fuel (SNF) which has become a challenge worldwide with South Africa also facing the same challenges.Radioactive waste management in South Africa is guided by a Radioactive Waste Management Policy and Strategy (RWMP&S) for the Republic of South Africa. In the Policy and Strategy, the government recognises that the practice of storing SNF in on-site storage pools is not sustainable and that regardless of any strategy chosen for the management of SNF, a long-term above-ground storage on an off-site facility is needed until a permanent solution is found.A number of options for dry storage systems are available from international vendors. The challenge is to select the most appropriate system for South Africa. The Analytic Hierarchy Process (AHP) is identified as a possible tool to aid the selection process. The study aims to demonstrate that the Analytic Hierarchy Process (AHP) is a tool that can be used to select a dry storage system option that will best meet South Africa’s needs.The study also presents the criteria to evaluate possible dry storage system options available in the international market. Pair-wise comparisons are conducted, and the ranking of the options is determined. Based on the results obtained from this study, it was demonstrated that the AHP tool can be applied to decide on the available option that will best meet South Africa’s needs.

Exploring the impact of temperature and oxygen partial pressure on the spent nuclear fuel oxidation during its dry management

The management of Spent Nuclear Fuel (SNF) comprises different stages in which security is demonstrated. Nevertheless, fundamental research can lead to other design options that must be considered. Currently, one of the focuses is the dry interim storage option, as the shortest-term solution until final repositories are available. During this stage, one concern is the oxidation of the fuel. If UO2 (SNF matrix) is exposed to air at high-enough temperature, formation of U3O8 takes place. The larger volume of this phase could entail stresses on the SNF clad, which is the first barrier to prevent radioactive material release. It is known that this oxidation is a temperature-dependent reaction and ensuring an inert atmosphere discards any effect during SNF dry management. However, at what extent temperature and oxygen concentration would have an impact on the U3O8 formation is not established, being the available experimental data very scarce. We follow this oxidation in representative ranges of temperature and oxygen concentration of dry storage facilities by using in-situ Raman spectroscopy. The results show that temperature is a more-affecting factor than the oxygen concentration at the studied conditions. Therefore, efforts to limit temperatures would yield more benefits in preserving fuel matrix integrity.

Spectroscopic and chemical investigations on volatile fission and activation products within the fuel-cladding interface of irradiated pressurised water reactor fuel rod segments

Abstract. In Germany, the present waste management concept foresees the direct disposal of spent nuclear fuel (SNF) in deep geological repositories for high-level waste available by 2050, at best. Until then, SNF is encapsulated in dual-purpose casks and stored in dry interim storage facilities. Licenses for both casks and facilities will expire after 40 years following loading of the cask and emplacement of the first cask in the storage location. Yet, due to considerable delays in the site selection process and the estimated duration for construction and commissioning of a final repository of at least 2 decades, a prolonged dry interim storage of SNF is inevitable (ESK, 2015). Concerning these considerable timespans, integrity of the cladding is of utmost importance regarding the ultimately conditioning of the fuel assemblies for final disposal. Various processes strain the structural integrity of Zircaloy cladding during reactor operation and beyond such as delayed hydride cracking, fuel-cladding chemical interactions or irradiation damage induced by α-emitters present in the fuel pellet's rim zone (Ewing, 2015). Especially with higher burn-up, the gap between fuel and cladding closes and results in the formation of an interaction layer, in which precipitates of fission and activation products are present, displaying an interface for degradation processes. For chemical analysis and speciation of these agglomerates, Zircaloy-4 and SNF specimens were sampled from fuel rod segments irradiated in commercial pressurised water reactors during the 1980s. Zircaloy-4 specimens were taken from an UOX (50.4 GWdtHM-1) and mixed oxide fuel (MOX) (38.0 GWdtHM-1). In addition, SNF fragments were sampled from the closed gap of both fuel types to examine volatile activation and fission products, which had been segregated from the centre to the pellet periphery during irradiation and thus contribute to the possible chemically assisted cladding degradation effect of the precipitates within the fuel-cladding interface. Spectroscopic analysis of precipitates within the interface layer between fuel and cladding were performed by optical microscopy, X-ray absorption and X-ray photoelectron spectroscopy, as well as by energy-dispersive scanning electron microscopy. Moreover, the radionuclide inventory of the respective Zircaloy-4, fuel and interaction layers was determined using liquid scintillation counting, γ-spectroscopy, gas mass spectrometry, ion chromatography and inductive-coupled plasma mass spectrometry and compared to results received by MCNP/CINDER and webKORIGEN calculations. In this study, we provide results regarding the speciation and chemical composition of previously identified Cs-U-O-Zr-Cl-I bearing compounds found in the interaction layer of irradiated nuclear fuel and inventory analyses of radionuclides present therein, with particular emphasis on Cl-36 and I-129. Furthermore, the agglomerates within the fuel-cladding interface were characterised for the first time utilising synchrotron radiation-based Cl K-edge and I K-edge measurements, resulting in compounds with structural similarities to CsCl and CsI. The outcomes obtained from this study provide further insights into the complex chemistry within the fuel-cladding interface with respect to the aging management and integrity of SNF under the conditions of interim storage. In future studies we will examine whether the different compounds at the fuel-cladding interface have the potential to affect the mechanical properties of Zircaloy cladding.

Safety Aspects in Dry Storage of Spent Nuclear Fuel in Long Term Operation for Brazilian Nuclear Power Plants

The use of Dry Casks Storage Systems (DCSS) as a strategy for Spent Nuclear Fuel (SNF) management is a solution widely used by several countries. It should be noted that this approach was originally conceived as a temporary / supplementary storage solution, since the ultimate solution is the reprocessing or the storage of SNF in Finals Repositories of radioactive material. Taking into account the limited storage capacity of Angra I and II Spent Fuel Pool, the Nuclear Power Plants Operator has submitted to the Brazilian Nuclear Regulatory Body (CNEN) the design of the Dry Storage Unit (UAS) as an alternative to SNF management. This work performs a Safety Evaluation applying the methodology defined in NUREG 1927 [2]. Three Structures, Systems and Components (SSC), classified as ITS (Important to safety), were selected. Following the methodology, materials and service environments were identified for those SSCs previously selected. In a second moment the ageing mechanisms and degradation effects to be considered in the UAS Long Term Operation were presented. As a result of this Safety Evaluation, some recommendations were extracted. These recommendations together with those found in the literature can guide the regulatory decision-making process regarding to the license renewal of UAS in Brazilian territory.

Assessment of activity resources by nomenclature of main radiation-hazardous radionuclides during the long-term management of spent nuclear fuel of the Belarusian NPP

The purpose of the study is calculation research of the radiation characteristics of fission products and actinides at different phases of spent nuclear fuel (SNF) management of the Belarusian Nuclear Power Plant (NPP). The study is aimed at the scientific support of the government decision as determined by the “On approval of the spent nuclear fuel management Strategy of the Belarusian nuclear power plant”. А probabilistic forecasting model and an effective code CUB for the spent nuclear fuel radioactivity inventory assessment were developed by the authors. Radionuclides activities as function of nuclear fuel burnup for nuclear fuel with the initial enrichment on the 235U equals to 4.81 % on the base of approximation relations of Regulation RB-093-14 (Moskow, 2014) have been calculated. Basic relations between specific activities of the main hazardous fission products and actinide, the SNF burnup and initial degree of fuel enrichment were analyzed. The rates of decrease of individual and total fission products and actinides activities of the Units №1 and 2 of the Belarusian NPP were obtained depending on the specific phase of spent SNF management. The results are of value for decision-making on ecology acceptable SNF management option introduced by Spent Nuclear Fuel Management Strategy of the Belarusian NPP.

Assessment of Permian Zubers as the Host Rock for Deep Geological Disposal

Together with renewable energy sources, nuclear power represents an important contribution to a sustainable energy mix in many countries and has an important impact on sustainable development. Nuclear energy production is also a source of high-level radioactive waste (HLW) and spent nuclear fuel (SNF), which require special concern. Disposal in deep geological formations is one of the solutions for the long-term management of HLW and SNF. It requires the development of a concept ensuring long-term safe isolation of waste and its validation applying the safety case methodology, which is a formal compilation of evidence, analyses and arguments that quantify and justify a claim that the repository will be safe. The results of laboratory testing of a potential repository host rock are an important component of the evidence that helps in the safety assessment of the deep geological disposal concept. This paper presents results of research focused on the physical, geomechanical and sorption properties of the Brown and Red Zuber unit rocks from the Kłodawa Salt Mine in Poland, which together with rock salt are an important component of Polish salt domes. The Brown and Red Zubers are typical evaporite lithostratigraphic units for the Polish part of the Zechstein Basin. They consist of halite (15–85%) and water-insoluble minerals, such as anhydrite, clay minerals, carbonates, quartz and feldspar, which occurred in varying proportions in the tested samples. The properties of the zuber rocks have been compared with those of rock salt, which is considered a suitable host rock for deep geological disposal of radioactive waste.

Corrosion tests in water of fuel elements irradiated in the world’s first NPP reactor

Spent nuclear fuel (SNF) storage in reactor spent fuel pools (SFP) is one of the crucial stages of SNF management technology: it requires special measures to ensure nuclear and radiation safety. During long-term storage in water-filled SFPs, leak-tight canisters in which SFAs are usually placed can become unsealed, which will result in the development of corrosion processes in the fuel element (FE) claddings. We studied fragments of spent fuel elements of the AM reactor of the World’s First NPP during their long exposure in the aqueous medium. The aim of the study was to obtain experimental data on the corrosion changes in the FE claddings and fuel composition during storage as well as on the release of radioactive fission products from them. For the study, a laboratory facility for exposing fuel elements in the water was developed and experimental fragments of fuel elements were made. The study was carried out in the hot chamber of the SSC RF-IPPE. The change in the activity of the water was estimated by the γ-dose rate from the selected water sample. The diameter measurements and metallographic studies were carried out in various sections of FE fragments. Corrosion tests were carried out on fragments of spent fuel elements of the AM reactor of the World’s First NPP that were stored for a long time (more than 50 years – FEs with U-Mo fuel and ~ 20 years – FEs with UO2 fuel) using standard technology – first in SFP canisters filled with water and then in dry canisters in the air. Placing the fuel elements in the water did not lead to through damage to the FE claddings and a significant change in the size (diameter) of the outer cladding. Metallographic studies of the FE fragments after the corrosion tests showed the presence of intergranular and local frontal corrosion on the surface of the claddings, the depth of which exceeded the depth of the cladding corrosion defects before testing. The rate of radionuclide release from the fuel composition was estimated by the γ-dose rate of water samples taken from the glasses with FE fragments. Throughout the test period, the dose rate of water samples from the glasses with defect-free FEs remained constant. The dose rate from water samples taken from the glasses with the FE fragments with an artificial defect grew during storage.

Analysis of the costs of spent nuclear fuel management in Spain: The Mariño model

In Spain, after almost a decade of delay in the construction of a centralised interim storage facility, the spent nuclear fuel (SNF) context has suffered a significant change. Thus, the aim of this article is to present the Mariño model, which analyses the costs of different back-end scenarios by means of the net present value and the material flows calculation, in order to re-examine the SNF management strategy for Spain and to help determine which alternative suits better this new context.The current strategy was analysed opposed to two direct disposal alternatives and a reprocessing option. The results show that the scenario of direct disposal without a centralised interim storage facility would decrease the current strategy costs about a 39%. Also, an alternative design for such facility could entail a 29% cost reduction. However, a reprocessing strategy could increase the costs about 140%, but if some assumptions of the scenario are changed, the costs could be significantly reduced. Additionally, an increase of the NPPs operational lifetime would decrease the levelized costs of all scenarios, which emphasises the economic advantage of longer operational lifetimes.

Management of Spent Nuclear Fuel from Research Reactors at the Kurchatov Institute

The problems of managing spent nuclear fuel stored in at-reactor repositories at the NRC Kurchatov Institute and current methods of inventorying the nuclear content of the repositories are discussed. Methods of identifying FA and leak-tight bottles with fuel rods by means of radiometric methods of measuring the radiation parameters of АF are discussed. The use of remote-controlled machines makes it possible to reduce the effect of high radiation fields on the workers and protect them in their work. The main problems arising during long-term storage of spent fuel are degradation of the protective barriers of the fuel rods and the lack of information on the radioactive content and the produced high-level wastes. Regular inventorying of the repositories ensures that the missing data are replaced, and in preparation for shipment attention is focused on restoring the outer casings of the packages for shipment to reprocessing facilities.

사용후핵연료 공론화위원회의 운영 개선 방안 : 프랑스 경험이 주는 교훈을 바탕으로

2013년 10월부터 609일간 시행한 ‘사용후핵연료 공론화’는 한국에서 처음으로 전 국민을 대상으로 시도한 공론화 사업이었다. 따라서 운영방식 면에서 완벽할 수 없었고 사후에 많은 논란도 낳았다. 더욱이 문재인 정부는 집권 후 이 공론화를 사실상 백지화시키고 사용후핵연료 공론화위원회를 재구성하여 다시 운영하기로 했다. 뿐만 아니라 공공갈등을 유발하는 대규모 국책 사업에 대한 사회적 수용성을 증진하기 위해서 향후 정부는 공론화 방식을 많이 활용할 가능성이 높다. 신고리 5·6호기 건설 재개 문제를 결정함에 있어서 이미 공론화를 시행하였고 그 결과에 따라 정책방향을 수정한 바 있다. 본 연구는 ‘사용후핵연료 공론화’ 운영방식에서 나타난 아래 문제점을 지적하고 개선 방향을 제시함으로써 향후 더욱 효과적인 공론화 체제를 정착시키는데 기여하고자 한다. 즉, 1) 공론화위원회의 독립성⋅중립성 문제, 2) 기구의 한시성과 법적 지위 문제, 3) “최종권고안” 문제, 그리고 4) 제도화 문제이다. 이에 대한 개선 방향은 공공정책 사업에 공론화를 20년 이상 운영한 경험이 있고 공론화 제도 역시 가장 잘 정비한 프랑스 모델에서 찾는다.

Radiation situation dynamics at the Andreeva Bay site for temporary storage of spent nuclear fuel and radioactive waste over the period 2002–2016

The Coastal Technical Base (CTB) №569 at Andreeva Bay was established in the early 1960s and intended for the refueling of nuclear submarine reactors and temporary storage of spent nuclear fuel (SNF) and radioactive waste (RW). In 2001, the base was transferred to the Russian Ministry for Atomic Energy and the site remediation began. The paper describes in detail the radiation situation change at the technical site in Andreeva Bay from 2002–2016, the period of preparation for the most critical phase of remedial work: removal of spent fuel assemblies. The analysis of aggregated indicators and data mining were used. The article suggests the best number and location of checkpoints needed to ensure sufficient accuracy of the radiation situation description. The fractal properties of the radiation field are studied using the Hurst index. The relationship between checkpoints was assessed using the method of searching for checkpoint communities. The decrease in the integral of the ambient dose equivalent rate (ADER) at the technical site was evaluated by the method of time series decomposition. Three components of time series were identified: trend, seasonal and residual. The trend of the ADER integral over the technical site is a monotonic decreasing function, where the initial and final values differ tenfold. Taking into account that 137Cs dominates the radiation situation on-site, it is clear that the ADER due to the radionuclide decay will have decreased by 1.4 times. It is estimated that only a small proportion of 137Cs has migrated off-site. Therefore, approximately a sevenfold decrease in dose rate is mainly due to remediation activities of personnel. During the year, the seasonal component varies the ADER integral by a factor of two, due to snowfall. The residual component reflects the uncertainty of the ADER integral calculation and phases of active SNF and RW management. The methods developed are used to support the optimization of remediation work as well as regulatory supervision of occupational radiation protection.

Innovative technologies for spent fuel safe management at Ignalina channel-type reactors

Abstract In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for “Channelized Large Power Reactor”-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

Регулирование обращения с отработанными ядерным топливом на пункте временного хранения в Губе Андреева на Кольском полуострове

Purpose: To share the experience in regulation of radiation safety and protection of workers involved in management of the spent nuclear fuel (SNF) and radioactive waste (RW), as well as radiation protection of the population and environment in the vicinity of sites for temporary storage at Andreeva Bay on the Kola Peninsula.
 Material and methods: Spent fuel from Russian nuclear powered submarines has been stored at shore based facilities for more than 20 years, notably at Andreeva Bay. The storage facilities were for some years poorly maintained and a significant fraction of the fuel that is still in store at the site wais damaged. Over the last years, hugemuch work has been done to improve the technical infrastructure and prepare for removal of the SNF from the temporary stores, management of existing RW. 
 Results: This paper presents progress with projects of the FMBA of Russia and NRPA cooperation for regulation of radiation safety and protection. During the researches, the following issues were addressed: radiological threat assessment to identify the priority directions of regulation; detailed analysis of the radiation situation on sites, at the territories and nearby the sites; radiation control and monitoring of the environmental conditions; development of the computer maps and geo-information system; emergency preparedness and response; improvement of radiation safety culture; etc. 
 Based on the received results of monitoring and assessment of the current risks, site-specific regulatory documents have been developed for the bodies and institutions under the FMBA of Russia involved in the activities to control the facility. Those documents include the requirements for: radiation protection of workers and population; personal dose monitoring; the RW management including the very low level RW; implementation of the environmental monitoring; radiation monitoring nearby the Andreeva Bay SevRAO facility; and remediation of the sites as remediation criteria and regulations. 
 The next stage of work is to carry out the regulation of large-scale removal of SNF during 2017–2021 and its subsequent transfer to Mayak PA, and operations to bring the infrastructure of the site into the hsafe conditions, i.e., ecological remediation of the site – by 2025. 
 Lessons learnt from this work are being used in support of improved international recommendations and guidance on how to address legacy issues.
 Conclusion: The experience accumulated during regulation of the remediation process of the former Naval Coastal Technical Bases, has helped to identify new relevant areas of improvement of the regulatory supervision at nuclear legacy sites. The study of potential hazard of radiation exposure to the personnel during technological operations of SNF and RW management is very important issue. In this light, the regulator in cooperation of the operator should develop some effective and efficient activities for dose monitoring. When dealing with the protection of the population and environment, a methodology of comprehensive radiation and chemical monitoring should be developed and models of radiation and chemical risks should be improved taking into account features of contamination of the site under remediation. An important link of the social focus of the regulator and the operator is to improve strategies of public communications near legacy sites under remediation.

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.